ML20054F672
| ML20054F672 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 06/14/1982 |
| From: | Cheng C, Hazelton W, Hinson C, Hum M, Danni Smith, Walton G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054F661 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8206170206 | |
| Download: ML20054F672 (29) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY C0tEISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of LONG ISLAND LIGHTING COMPANY
)
DocketNo.50-322(OL)
(Shoreham Nuclear Power Station, Unit 1)
)
NRC STAFF TESTIMONY OF WARREN S. HAZELTON, C. Y. CHENG, MARTIN R. HUM, DAVID E. SMITH, GLENN A. WALTON, AND CHARLES S. HINSON ON PRESERVICE AND INSERVICE INSPECTION PROGRAMS (SCContention25)
(SOCContention19(a))
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0322 PDR T
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i OUTLINE OF TESTIMONY Suffolk County's Contention 25 and Shoreham Opponents Coalition's Contention 19(a) allege various inadequacies in the Preservice (PSI) and Inservice (ISI)inspectionprogramsatShoreham. The testimony addresses the Staff review of the Shoreham PSI and the guidelines that the Staff will follow in its forthcoming review.of the Shoreham ISI.
ThetestimonyalsoaddressesthespecificallegationsinSOC19(a)and explains why these allegations are unfounded.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATDMIC SAFETY AND LICENSING BOARD In the Matter of LONG ISLAND LIGHTING COMPANY DocketNo.50-322(OL)
(Shoreham Nuclear Power Station, Unit 1)
NRC STAFF TESTIMONY OF WARREN S. HAZELTON, C. Y. CHENG, MARTIN R. HUM, DAVID E. SMITH, GLENN A. WALTON, AND CHARLES S. HIN50N ON SC CONTENTION 25 AND SCC CONTENTION 19(a)
Q.
Please state your names and positions with the NRC?
A.
(WSH) My name is Warren S. Hazelton.
I am the Section Leader of the Materials Application Section of the Materials Engineering Branch, Division of Engineering. A copy of my professional qualifications is attached.
(CYC) My name is C. Y. Cheng.
I am the Section Leader of the Inservice Inspection Section of the Materials Engineering Branch, I
Division of Engineering. A copy of my professional qualifications is attached.
(MRH) My name is Martin R. Hum.
I am a senior materials engineer assigned to the Materials Engineering Branch, Inservice Inspection Section, Division of Engineering. A copy of my professional l
qualifications is attached.
(DES) My name is David E. Smith.
I am a senior materials engineer assigned to the Materials Engineering Branch, M,aterials Application Section, Division of Engineering. A copy of my professional qualifications is attached.
w ap e--w (GAW) My name is Glenn A. Walton.
I am a senior resident inspector assigned to the NRC's Region I office. A copy of my professional qualifications is attached.
(CSH) My name is Charles S. Hinson.
I have been a health physicist with the Radiological Assessment Branch in the Office of Nuclear Reactor Regulation of the NRC since 1976. A copy of my professional qualifications is attached.
Q.
What is the purpose of this testimony?
A.
(ALL) The purpose of this testimony is to address SC Contention 25andSOCContention19(a)whichstate:
SC 25: ASME SECTION XI (PSI /ISI) PROGRAM Suffolk County contends that LILC0 has not adequately demonstrated the effectiveness of the technology and methods available that are required to satisfy the inspection and tests specified b'y 10 C.F.R. 50, Appendix A, GDC 32, 36, 39, and 45.
The technology used for the PSI inspection for the reactor pressure boundary cannot be correlated to that used for the ISI program. And, further, the results from inspected areas of the reactor pressure boundary cannot be extended to non-inspected areas. Suffolk County further contends that the Shoreham plant does not comply with 10 C.F.R. 50.55a(g) which requires, for the ISI Program, use of the Edition and Addenda of Section XI of the ASME Code in effect 12 months prior to the date of issuance of the operating license. Because the Shoreham piping configuration and reactor vessel design substantially pre-date the latest code, LILC0 has already identified some Section XI inspection requirements for which exemption has been requested. Additional ex-emptions and/or waivers will undoubtedly be identified. The impact of these deficiencies has not been specified and analysis has not been presented to demonstrate the effectiveness of the ISI program.
. S0C19(a): REG. GUIDES 1.2 AND 1.150: RPV INTEGRITY AND TESTING A major contributing factor in the TMI-2 accident was that operating plants were not required by the NRC Staff (Staff) to be in compliance with current regulatory practices (i.e., Regulatory Guides, Branch Technical Positions, and Standard Review Plans). The TMI-2 accident also demonstrated that the current regulatory practices, practices similar to those being applied by the Staff in their safety evaluation of Shoreham, were in a number of cases not suitably conservative to prop (erly protect the health and safety of the public i.e., hydrogen-generation, radiation shielding, source terms, and single failure criterion).
S0C contends that the NRC Staff has not required LILC0 to incorporate measures to assure that Shoreham conforms with the standards or goals of safety criteria contained in recent regulatory guides. As a result, the Staff has not required that Shoreham structures, systems, and components be backfit as required by 10 C.F.R. 5 50.55a, 9 50.57, and 6 50.109 with regard to:
(a) Regulatory Guides 1.2 and 1.150. -- LILC0 has not adequately demonstrated that the design, preservice examination, and inservice ex-i l
amination of the Shoreham reactor pressure l
vessel and vessel nozzles is in compliance with the requirements of 10 C.F.R. Part 50, Appendix A, Criteria 1 and 31,10 C.F.R.
i 6 50.55a, and 10 C.F.R. Part 50, Appendix B,
[
Criteria XII and XVII, in that:
1.
Quality control of the ultrasonic testing (UT) equipment including the UT transducers does not meet Regulatory Guide 1.150 and thus is l
inadequate to provide reliable and repro-ducible UT results.
2.
UT examination 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.
3.
Radiation exposure to examination personnel has not been demonstrated as meeting ALARA.
4.
Structural integrity of the pressure vessel has not been demonstrated in accordance with Regulatory Guide 1.2 and thus there is
inadequate assurance that failure of the vessel by brittle fracture as a result of the design basis accident will not occur.
Q.
How is your testimony divided?
A.
(ALL) The contentions raise a number of issues. Mr. Hazelton, Dr. Cheng, Mr. Hum and Mr. Walton will address the subject of nondestructive examination; Mr. Hazelton, Dr. Cheng and Mr. Smith will address fracture toughness properties; and Mr. Hinson will address ALARA.
Q.
What regulation governs the preservice and inservice in-spections at Shoreham?
A.
(CYC,MRH) Paragraph 50.55a(g) of 10 C.F.R. Part 50 establishes the requirements for the preservice and inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 components.
The basic requirements for the preservice inspection are defined in subparagraph 50.55a(g)(2)whichstates,inpart:
For a boiling or pressurized water-ccoled nuclear power facility whose construction permit was issued on or after January 1,1971, but before July 1, 1974, components (inluding supports) which are classified as ASME Code Class 1 and Class 2 shall be designed and be provided with access to enable the performance of (1) inservice examination of suchcomponents(includingsupports)and(ii) tests for operational readiness of pumps and valves, and shall meet the preservice examination requirements set forth in editions of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda in effect 6 months prior to the date of issuance of the construction permit. Thecomponents(including supports) may meet the requirements set forth in subsequent editions of this code and addenda which are incorporated by reference in paragraph (b) of this section, subject to the limitations and modifications listed therein.
The pertinent requirements for the inservice inspections are defined insubparagraph50.E5a(g)(4)whichstates,inpart:
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Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) which are clas-sified as ASME Code Class 1, Class 2 and Class 3 shall meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda that become effective subsequent to editionsspecifiedinparagraphs(g)(2)and(g)(3) of this section and are incorporated by reference inparagraph(b)ofthissection,totheextent practical within the limitations of design, geometry and materials of construction of the components.
(1)
Inservice examinations of components, in-service tests to verify operational readiness of pumps and valves whose function is required for safety, and system pressure tests, conducted during the initial 120-month inspection interval shall comply with the requirements in the latest edition and addenda of the Code incorporated by reference in paragraph (b) of this section on the date 12 months prior to the date of issuance of the operating license, subject to the limitations and modifications listed in paragraph (b) of this section.
Q.
What are the applicable editions of the ASME Code for the preservice and inservice inspections at Shoreham?
A.
(MRH) Shoreham received its construction permit on April 14, 1973. The regulation requires compliance with the preservice examination requirements in effect six months prior to the date of issuance of the construction permit.
For Shoreham, the applicable ASME Code would have been the 1971 Edition including Addenda thru Winter 1971 (71E 71W) of Section XI. However, the Applicant elected for Shoreham to optionally update the preservice inspection program to meet the requirements of the 1971 Edition including Addenda thru Sumer 1972 (71E 72S) of Section XI, as permitted by the regulation.
l Q.
Describe the responsibilities of the various NRC organizations involved with the preservice and inservice inspections.
A.
(WSH, CYC, MRH, GAW) During construction, the Regional Offices review the Applicant's preservice examination program, selectively review examination procedures, and audit and witness various examination results, both at fabrication shops and plant site. The Regional Offices may request that an Applicant repeat. specific examinations. Audits are performed by the NRC to assess compliance with the requirements of the preservice in-spection program.
The Office of Nuclear Reactor Regulation also evaluates the pre-service and inservice inspection programs as part of its operating license review. The Materials Engineering Branch (MTEB) is respnsible for the review of the nondestructive examination of ASME Code Class 1, 2, and 3 components, including the piping system, and the system hydrostatic tests. The Mechanical Engineering Branch is responsible for the review of the testing of pumps and valves.
The MTEB review of the preservice inspection (" PSI") program consists of two basic areas:
(1)determiningwhetherthemethodof examination and selection of the components and piping system welds to be examined are in accordance with the applicable ASME Code Section XI and regulatory requirements and (2) evaluating ASME Code Section XI re-quirements that the Applicant determines to be impractical to perform.
This review of the PSI Program is completed before an operating license is issued. MTES also reviews the inservice inspection ("ISI") program after an operating license is issued as a condition to the license.
.2 r.
In addition, other organizations within NRC may have lead responsibility concerning specific generic issues that may require additional augmented examinations of certain components.
Q.
Describe the extent of the preservice examination at Shoreham?
A.
(CYC,MRH) The Applicant is performing a preservice examination of the reactor coolant pressure boundary (RCPB) based on (71E 72S) of Section XI of the ASME Code. The RCPB consists of all the ASME Code Class 1 components and certain ASME Code Class 2 components.
In addition, the Applicant is performing examinations on a representative sample of ASME Code Class 2 components that are not part of the RCPB.
Q.
Suffolk County contends in SC 25 that LILCO has not adequately demonstrated the effectiveness of the technology and methods available that are required to satisfy the inspection and tests specified by 10 C.F.R. 50, Appendix A, GDC 32, 36, 39 and 45. Suffolk County further contends that the technology used for the PSI inspection for the reactor i
pressJre boundary cannot be correlated to that used for the ISI Program.
Could you please address these allegations?
A.
(WSH,CYC,MRH) For the components subject to examination in accordance with GDC 39, " Inspection of Containment Heat Removal System,"
and GDC 45, " Inspection of ' Cooling Water Systems,"Section XI generally requires either a surface examination, visual examination under system pressure, visual examination, or hydrostatic test. The technology required to perform the above stated methods of examination is well established.
Only GDC 32," Inspection of Reactor Coolant Pressure Boundary," and GDC 36,
" Inspection of the Emergency Core Cooling System," relate to the PSI of l
the RCPB. For components subject to examination in accordance with
GDC 32 and GDC 36,Section XI generally requires either volumetric (ultrasonic or radiographic) examination methods or the methods used to comply with GDC 39 or GDC 45.
Direct correlation of PSI examination results for use in the ISI examination is not always necessary. Each inspection performed on a specific weld stands on its own merit.
In the event that a flaw in-dication is detected during the ISI examination with a different or improved technique, this flaw indication must be considered a new indication unless detected during preoperational examinations and dis-positioned in accordance with the examination acceptance standards defined in Section XI.
Q.
What document is LILC0 using to perform the preservice examination?
A.
(MRH) LILC0 is performing the preservice inspection based on the document entitled "Preservice Inspection Program Plan." Revision 3 of this document is available in the Public Document Room.
Q.
Do the nondestructive examination (NDE) procedures in the "Preservice Inspection Program Plan" meet the requirements of the applicable editions of Section XI?
A.
(GAW) During the period from 1975 through 1981, the NRC Staff performed a detailed review of twelve (12) NDE procedures in the Shoreham Preservice Inspection Program Plan. Certain modifications to the procedures were required. Based on this audit, we are satisfied that the Shoreham procedures meet the requirements of the 1971 Edition including Addenda through Summer 1972 of Section XI of the ASME Code.
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Q.
Does the Staff have any reason to believe the Shoreham pre-service inspection procedures might not be effective?
A.
(GAW) No. The effectiveness of specific procedures is based on compliance with Section XI requirements referenced in 10 C.F.R. 50.55a(b). All nondestructive test procedures must be qualified by demonstrating their effectiveness to the satisfaction of the " Authorized Inspector," an independent official whose responsibilities are defined in Section XI.
In addition, the NRC Staff has performed detailed audits and witnessed tests, including calibration sensitivity verification and determined these practices meet the code requirements.
Q.
Was the PSI data documented in a manner to permit correlation with the ISI results?
A.
(GAW) Yes. The NRC Staff review indicated that both the manual and automated ultrasonic testing records are being documented based on Article IS-600 " Records" of Section XI for future comparison with inservice examinations.
Q.
Suffolk County also contends in SC 25 that the results from inspected areas of the RCPB cannot be extended to non-inspectable areas.
Suffolk County further contends that the Shoreham plant does not comply l
^
with 10 C.F.R. 50.55a(g) which requires, for the ISI Program, use of the Edition and Addenda of Section XI of the ASME Code in effect 12 months prior to the date of issuance of the operating license. Could you please comment on these alla.gations?
A.
(WSH,CYC,MRH)Section XI inservice inspection requirements are based on examining a representative sample of welds to detect
. potential generic degradation. All welds will generally not be examined during the inservice inspections. A utility normally uses the results from the preservice examination to select the representative sample from areas without design restrictions to the extent practical.
The effective Edition of Section XI of the ASME Code for ISI cannot be identified because the date of issuance of the operating license is not known. A condition to the operating license will require that an acceptable ISI Program be submitted for review after licensing.
Q.
Suffolk County further contends that because the Shoreham piping configuration and reactor vessel design substantially pre-date the latest Code, LILC0 has already identified some Section XI inspection requirements for which exemption has been requested; additional ex-emptions and/or waivers will undoubtedly be identified; the impact of these deficiencies has not been specified; and an analysis has not been presented to demonstrate the effectiveness of the ISI Program.
Please address these allegations?
A.
(WSH,CYC,MRH) The preservice examination at Shoreham is still being conducted. The Applicant has indicated that some ASME Code examinations have been determined to be impractical. However, the Applicant has not identified the impractical requirements on the docket with a supporting technical justification. We will evaluate the Applicant's request for relief from impractical examination requirements and proposed alternative requirements and report our conclusions in a Supplement to the Safety Lyaluation Report.
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Q.
WearenowgoingtoaddresstheallegationsinSOC19(a). What are the specific examination requirements for the reactor vessel and i
vessel nozzles?
A.
(MRH)
(1) Preservice Examination - The (71E 72S) of Section XI requires a preoperational examination of essentially 100 percent of the pressure-containing welds in the reactor coolant pressure boundary (RCPB), which includes the entire reactor vessel and nozzles.
(2)
Inservice Examination (a) The (71E 725) of Section XI requires an inservice inspection that covers at least 10 percent of the length of each longitudinal weld and 5 percent of the length of each circum-ferential weld in the vessel shell during each 10 year inspection interval.
(b) The 1980 Edition of Section XI, for example, requires an inservice inspection that covers essentially 100 percent of the weld length of each longitudinal and each circumferential weld in the vessel shell during the first 10-year inspection interval.
Q.
Describe the preservice examination of the RPV.
l A.
(WSH,CYC,MRH,GAW) A manual ultrasonic examination was performed on the vessel shell welds and nozzle-to-vessel welds in the fabrication shop.
Rejectable flaws were detected and repaired in the fabrication shop. A manual ultrasonic examination was also performed at the plant site before installation. The Applicant also performed a confinnatory ultrasonic examination with mechanized equipment.
In addition, a suspect area was found in the reactor pressure vessel nozzle N4B. The suspect area was recently reported acceptable to the
requirements of Section XI, however, the NRC will review this data after a new calibration block is made, the area retested, and the results evaluated.
Q.
Does the existing design of the Shoreham 2PV permit an inservice inspection that ;neets the current requirements of Section XI of the ASME Code?
A.
(WSH,CYC,MRH,GAW) The Applicant has not identified the specific limitations to inservice inspection because the Code in effect twelve months prior to the issuance of the operating license cannot be defined. Based on the review of the layout of the fixed tracks for mechanized examination, the Applicant probably cannot examine 100 percent of the vessel shell welds.
Q.
When will the review of limitations to examinations be completed?
A.
(WSH,CYC,MRH) The Applicant has coninitted to identify all limitations to the preservice examination requirements, with a supporting technical justification, before the issuance of an operating license.
The NRC Staff will complete the evaluation on the Shoreham Inservice Inspection Program as a condition to the license before the first re-fueling outage when inservice inspections will be performed.
Q.
What is the NRC position concerning the implementation of Regulatory Guide 1.150 as related to Shoreham?
A.
(WSH,CYC,MRH,GA'.!) Regulatory Guide 1.150 is entitled
" Ultrasonic Testing of Reactor Vessel Welds During Preservice And Inservice Examinations." To the best of our knowledge, no applicant or licensee has received written instructions to implement this Regulatory Guide although some utilities have voluntarily incorporated the requirements of Regulatory Guide 1.150 into their reactor vessel
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inspections.
In addition, there is a distinct possibility that a revision to Regulatory Guide 1.150 will be proposed by the NRC before implementation.
According to the current version of Regulatory Guide 1.150, the effective date for preservice examinations to comply with the Regulatory Guide is January 15, 1982. The manual ultrasonic inspactions in the fabrication shop and the plant site of the Shoreham RPV were performed before that date.
Q.
. hat were the major objectives of Regulatory Guide 1.150?
W A.
(WSH,CYC,MRH) Regulatory Guide 1.150 describes provisions to augment the ultrasonic testing requirements defined in Section XI of the ASME Code. The major objectives cf Regulatory Guide 1.150 were to improve the capability to detect surface or near surface flaws in the examination region nearest to the ultrasonic transducer and to require the documentation of specific limitations to examination. Regulatory Guide 1.150 was directed primarily towards the sizing of detected in-dications on the inside surface of PWR vessels as related to the issue of pressurized thennal shock.
PWR plants typically perform the RPV ex-amination with mechanized devices from the inside of the vessel after removal of the internals. BWR plants typically perform the RPV ex-amination from the outside of the vessel.
Q.
What is the NRC position concerning the implementation of the ASME Section XI requirements for inservice inspection?
A.
(WSH,CYC,MRH) The Section XI inservice examination require-ments were originally established to detect generic degradation rather than defects in each individual weld as reflected in the (71E 725) sampling concept. Current Section XI sampling concepts require that essentially 100 percent of the length of each longitudinal and
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circumferential weld in the vessel shell be examined. Many PWR plants can approach the current Section XI requirements to examine essentially 100 percent of the vessel shell welds. However, many BWR plants, in-cluding Shoreham, cannot meet this current requirement. Therefore, the examination concept for BWR plants is still to examine for potential generic degradation rather than defects in the entire length of each weld. Our licensing basis for the inservice inspection of BWR reactor pressure vessels to assure safe operation is the examination of all inspectable welds to detect generic degradation, the system hydrostatic pressure tests, the original design, the fabrication examination, and the fracture toughness properties of the material.
Q.
S0C Contention 19(a)(1) alleges that the quality control of the ultrasonic testing (UT) equipment including the UT transducers does not meet the provisions of Regulatory Guide 1.150. Could you please address this allegation?
A.
(WSH,CYC,MRH)Section XI does not have requirements regarding the use of specific UT instrumentation.Section XI requirements are based on performance parameters and calibration confinnation of the entire examination system. The UT transducer provisions in Regulatory Guide 1.150 are intended to obtain supplemental information in the event that flaws are detected and require independent evaluation by the NRC.
Five provisions for the quality control of the UT equipment are contained in the Guide's regulatory position section entitled " Instrument Performance Checks." One provision deals with the frequency of perfonnance checks and the others deal with the recording and sizing of reflectors. Our conclusion regarding these quality control provisions related to the Shoreham RPV preservice inspections is that they are not technically
significant. During the fabrication shop and plant site preservice inspections, flaws that are significant to the safe operation of the reactor vessel have already been detected regardless of the additional performance checks described in the Regulatory Guide.
Q.
What is the technical significance of the UT examination travel timedescribedinS0CContention19(a)(2)?
A.
(WSH,CYC,MRH,GAW)Section XI allows static calibration on the calibration standard.
Indications would be recorded automatically while the transducer is in motion. The intent of the Regulatory Guide is to assure adequate sensitivity is obtained while calibrating by requiring the tranducer be traversed at scanning speeds over the calibration holes while calibration is perfonned. Specifically, Regulatory Guide 1.150 Section C(6) states that:
Indications that travel on the horizontal baseline of the scope for a distance greater than indi-cations from the calibration holes (at 20 percent DAC amplitude) should be recorded.
Indications that travel should be recorded and sized at 20 percent DAC.
The LILC0/NES procedure 80A0462 " Manual Ultrasonic Examination Procedure For Reactor Pressure Vessel Circumferential and Longitudinal Welds,"
states in paragraph 11.1.1 that:
All indications which exceed 6 dB below DAC (50% of the reference level) shall be recorded on the appropriate data sheet at the time of the weld examination and prior to removing equipment from the reactor pressure vessel.
In addition, paragraph 11.2.1 states that:
All recordable (11.1.1) indications which exceed DAC or are interpreted to be cracks, lack of penetration, or lack of fusion as per the re-quirement of ASME Boiler and Pressure Vessel Code,Section XI, IS-300, shall be reported to the client by the end of the shift.
. l The above defined indications (cracks, lack of penetration or lack of fusion), are all considered rejectable indications regardless of size per the fabrications and preservice inspection code requirements. Our conclusion regarding the significance of the UT examination travel time is that the requirements of LILC0/NES procedure 80A0462, paragraphs 11.1.1 and 11.2.1 will act.omplish the objective of the Regulatory Guide of detecting and reporting flaws that are significant to the safe operation of the reactor vessel.
Q.
S0CContention19(a)(3)allegesthatradiationexposureto examination personnel has not been demonstrated as meeting ALARA. Could you please comment?
A.
(CSH) The restrictions governing entry into a controlled area (area controlled for the purpose of protecting individuals from exposure to radiation) are the same for examination personnel as they are for plant personnel. Prior to performing any work involving entry into Radiation Areas, High Radiation Areas, contaminated areas, areas exposed to neutron radiation, or areas having airborne radioactivity, all persons must receive authorization through specific Radiation Work Permits (RWP) issued by HP Section personnel. These permits state radiation levels in the area, allowable stay times, protective clothing and respiratory protective equipment required, monitoring requirements, and any special notes or cautions pertinent to the specific job (such as additional shielding requirements). RWP's also specify any special tools or equipment which will be required for the job. These permits assure that all work is perfonned in a radiologically saf e manner and that doses are maintained ALARA.
In addition to having a RWP, all personnel entering a controlled area must have satisfactorily completed a level of Health
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. Physics Training appropriate to the controlled area in question. Direct Reading Dosimeters, as well as TLD's, will be assigned to all personnel entering controlled areas. These will be used to keep track of an individual's dose to ensure that he does not exceed the Administrative Dose Control limits.
It is in the above ways that occupational radiation exposure to plant and contractor personnel, including examination personnel, are maintained as low as is reasonably achievable.
Q.
In S0C 19(a)(4), S0C contends that the structural integrity of the pressure vessel has not been demonstrated in accordance with Regulatory Guide 1.2, and thus there is inadequate assurance that failure of the vessel by brittle fracture as a result of the design basis accident will not occur. Could you please comment?
A.
(WSH,CYC, DES) Regulatory Guide 1.2 is entitled " Thermal Shock to Reactor Pressure Vessels." The thermal shock event is presently under extensive review within the NRC. The NRC Staff does not believe BWRs have a significant Pressurized Thermal Shock (PTS) concern, for several reasons. Most importantly, BWRs operate with a large portion of the water inventory inside the pressure vessel at saturated conditions, (that is, it exists as a mixture of steam and liquid water at the mixture's boiling temperature and pressure). Any sudden cooling will condense steam and result in a pressure decrease, so simultaneous creation of high pressure and low temperature (necessary to cause a PTS concern) is very improbable. BWR operating experience provides verification that PTS events are very improbable since there have been no significant PTS events at any domestic or foreign BWR (that is, significant pres-surization during or after a severe overcooling has not occurred).
f
. Also contributing to the lack of PTS concerns for BWRs is the lower fluence of the vessel inner wall, since BWRs have more water between the core and the vessel wall due to the recirculation flow path (water shields the vessel from the core). Finally, the operating pressure of l
BWRs is lower, which results in a lower stress intensity at the bottom of I
I a postulated crack.
The fracture toughness data indicate that the structural integrity of the reactor pressure vessel is in accordance with Regulatory Guide 1.2.
Adequate assurance has been demonstrated to the Staff that failure of the vessel by brittle fracture as a result of the design basis accident will not occur.
Coul' you please provide a brief conclusion to your testimony?
d Q.
A.
(ALL) Our review indicates that LILCO has adequately provided for meeting the inspection requirements of Section XI of the ASME code and 10 C.F.R. 9 50.55a(g). Our review also indicates that the inspection work will be performed in a radiologically safe manner and that doses will be maintained ALARA. The fracture toughness data indicate that the structural integrity of the reactor pressure vessel is in accordance with Regulatory Guide 1.2.
Therefore, we conclude that the allegations are without foundation.
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WARREN 5. HAZELTON h j PROFESSIONAL QUALIFICATIONS f
4 My name is Warren S. Hazelton. In my capacity as Acting Chief of the s
Materials Engineering Branch and Section Leader, Engineering Materials Application Section, in the Division of Engineering, NRR, I am responsible for reviewing materials related aspects of operating nuclear power plants and Safety Evaluation Reports for licensing new plants. In conjunction with this work, I am also responsible for aiding in the preparation of Federsl Regulations and Regulatory Guides relating to materials, inservice in-spection, and operational limitations important to the safety of nuclear power plants. Another primary responsibility is reviewing research programs on reactor safety, evaluating results of these programs, making recomendations for new programs, and factoring the results of these programs into our other review activities.
I was born in Cutler, Minnesota on October 20, 1916, and attended 5
public schools in Duluth and Wahkon Minnesota. After attending the University of flinnesota intermittently, I joined the Armed Services in 1941.
I was discharged in 1945 after serving as an Arv Air Force Pilot.
I then resumed y education at the University of Minnesota, was honored 1
by being selected for " Plumb Bob," and graduated in 1949 with a Bachelor of Metallurgical Engineering degree, with distinction.
From 1949 to 1960. I was employed in the Westinghouse Aviation Gas Turbine Division, at South Philadelphia and at Kansas City, Missouri.
From 1954 to 1960. I was manager of the materials application and develop-ment activity, responsible for the materials aspect of design, materials properties, failure analysis, and the development of new materials.
From 1960 to 1963 I was Supervising Engineer of the Materials Develop-ment Section at the Westinghouse Bettis Atomic Power Laboratory.
In this capacity I was responsible for development programs in the fields of stress corrosion, brittle fracture prevention, and radiation damage.
From 1963 until 1972, when I assumed g present position. I held various management positions in the Westinghouse PWR Systems Division. My respolibilities included the development and application of improved fracttre prevention technology, evaluation of radiation damage, stress corrosion prevention, and involved close interface with design groups.
I was responsible for the detailed failure analysis performed on the inter-nals at the Yanke6 Rowe, Connecticut Yankee Trino (Italy), and SENA (Franco-Belg.) plants.
I also participated actively in the redesign and repair work performed for these plants.
I have been active in the preparation of Codes and Standards relating to reactor safety.
Specifically, I am a member of several ASME Boiler and Pressure Vessel Code committees, the Pressure Vessel Research Committee Task Group on Fracture Toughness Requirements, and several ASTM committees developing standards for evaluating radi'ation damage of metals.
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._ _._ _ _ c ATTACHMENT PROFESSIONAL QUALIFICATIONS OF C. Y. Cheng EXPERIENCE: January 1982 Section Leader to Date Inservice Inspection Section Materials Engineering Branch Division of Engineering As a Section Leader, I am responsible for providing technical supervision and direction to a group of materials engineers conducting reviews and evaluation of the inservice inspection aspects of the reactor coolant pressure boundary and safety-related systems as described in the appli-cations for Construction Permits and Operating Licenses of nuclear power plants and in the proposed amendments to operating licenses.
December 1973 Materials Engineer /
to December 1981 Principal Materials Engineer Div. Tech Review /Div.
Operating Reactors /
Div. Licensing, USNRC Served as a principal reviewer for material engineering aspects of oper-rating reactor problems and issues related to plants under construction.
August 1967 Research Metallurgist to December 1973 Materials Science Division Argonne National Laboratory Argonne, Illinois Performed basic and applied researches on the mechanical behavior of metals and alloys.
September 1963 Research Assistant /
to August 1967 Postdoctoral Research Metallurgist Lawrence Radiation Lab.
University of California i
Berkeley, California Conducted research on the mechanical metallurgy of alloys.
l September 1961 Research Assistant to August 1963 Denver Research Institute University of Devener 7
Denver, Colorado i
Conducted research on the lithium - rhodium - hydrogen systems.
February 1960 Full time Teaching Assistant to August 1961 Mechanical Engineering Dep.
National Taiwan University Taipei, Taiwan
2-Served as a full time teaching assistant for " Engineering Materials" course and its lab. Also conducted research on the mechancial properties of ferritic steels and Al-Si alloy EDUCATION:
PhD in Engineering Science (Metallurgy), University of California, Berkeley, California, MS in Metallurgy, University of Denver, Denver, Colorado BS in Mechanical Engineering, National Taiwan University, Taipei, Taiwan.
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e ATTACHMENT PROFESSIONAL QUALIFICATIONS 0F s
- Mar. tin R. '. Hum EXPERIENCE: December 1974 Materials Engineer to Materials Engineering Branch Date Division of Engineerins'
- I am currently a senior materials engineer in the Materials Engineering Branch, Division of Engineering.
% responsibilities include the review of inspection criteria for structural and mechanical components.
I have participated as a technical reviewer in evaluating nondestructive -
examination for applications for construction pemits and operating licenses for power reactors and DOE,-owned operating faciliti.gs exempt
.from the licensing process.
My specific asignments include review of operating license ahications for compliance with Standard Review Plans for which the Insenice Inspection Section is responsible.
June 1965 Mechanical Engin'e'eN to U.S. Army Facilities Engineering Support December 1974 Agency Fort Belvoir, Virginia i
Prior to joining the NRC, I was employed as a civilian mechanical engineer a
with the U.S. Army Facilities Engineering Support Agency.
% responsi-bilities included the evaluation and implementation insenice inspection for structural and mechanical components in mobile nuclear power plants.
I have participated as a project engineer during the inspection, modifi-cation and/or repair of nuclear power plant systems.
l EDUCATION:
Master of Science (Mechanical Engineering), George Washington University, 1971.
5achelor of Mechanical Engineering, George Washington University,1965.
PROFESSIONAL LICENSE:
I am registered to practice Profession ^al Engineering in the District of Columbia, License Number 6351.
PROFESSIONAL SOCIETY MEMBERSHIP:
I am a member of the American Society of Mechanical Engineers.
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s ATTACHMENT PROFESSIONAL QUALIFICATIONS DF.
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- David E. Smith EXPERIENCE: March 1980 Senior Materials Engineer to Materials Engineering, Branch Date Division of Engineering Knowledgeable and experienced in welding, fabrication and
-lnspection of materials and other related engineering aspects of nuclear reactors.
Serves as a qualified materials engineer in the Materials Engineering Branch, Division of Engineering.
Responsible for reviews, analyses, and evaluation of safety issues related to structural and mechanical components of reactor facilities licensed for power operation.
Participates as a technical reviewer in evaluating applicatipns for construction permits and operating licenses for power and.non-power reactors and operational and design modifications of DOE arid.D0D-owned operating facilities exempt from the licensing process.
Specific as(ignments include' review of operating license aplications for compliance with Standard Review Plans for which the.1$aterials
' Application Section is responsible.
EDUCATION:
Bachelor of Metallurgical Engineering, Rensselaer Polytechnic Institute,1959 EXPERIENCE:
(Prior to joining NRC)
May 1967 to Materials Engineer Naval Sea Systems Command, Code 05E2, March 80 Washington, D.C.
Responsible for materials specifications, Hull material development programs, consultant on welding, fabrication and inspection of metal structures, material selection, corrosion, machinery materials problems.
The hull material development programs involved basic alloy research, the t
l making and processing of all structural metal forms (castings, forgings, plate, extrusions, weld wire, rolled' product), their fabrication (welding, cutting, machining, forming, painting), structural tolerances,- and evalu-ation of structural performance, strength, toughness, corrosion, fatigue, compatibility with other materials, etc.
I would interface with material manufacturers, suppliers, shipyards, and designers, and the type desks responsible for providing ships to the fleet.
April 66 to Student.
Acquired commercial and instrument ratings for May 67 single engine land airplanes.
Dec 64 to Manufacturing Engineer for Ling Temco Vought, Centerline, April 66 MI.
Developed welding procedures for the LANCE missile tankage assembly.
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GLENN A. WALTON EKFERIENCE:
August 3,1981 to Date - Senior Resident Inspector Beaver Valley, Unit 2 NRC: Region I As senior resident inspector at Beaver Valley, Unit 2. I perform inspection of the construction site to ascertain compliance with the construction permit and NRC rules and regulations.
March 11, 1974-to August 3,1981 - Reactor Inspector, Reactor Construction and Engineering Support Branch., NRC: Region I Responsitle for inspaction of nuclear power plants during construction and operation phases in specialty areas of nondestructive examination, fabrication of pressure -
vessels and fabrication of piping systems.
Principal inspector of Pre-Service and In-Service Inspection activities at nuclear power plants.
1971 to 1974 - Quality Control Manager, Babcock & Wilcox Co., Mt. Vernon, Indiana Respcnsible for all aspects of Quality Control at the B&W heavy pressure vessel shops.
Fabrication inc1'aded reactor vessels, steam generators, pressurizers and primary piping for comercial nuclear' application and reactor vessels and heads for Navy nuclear applications.
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~1968 to 19/1 - Nontestructive Test Section Head, Babcock & Wilcox Co., Mt. Vernon, Indiana Responsible for all aspects of nondestructive examinations on nuclear components.
'As the designdted qualified Level III in accordance with SNT-TC-1A. Successfully pr.ssed the NAVSHIpS 250-1500-1 requiregnt for ultrasonic test examiner, certification number 0086.
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1963 to 19G8 - Nondestructive Test Supervisor, Babcock & Wilcox Co., Barberton, Ohio
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Directed technicians in performance of ultrasonic, magnetic particle, and liquid penetrant czaminations.
QualifiedinMT,PTandUTasLevelIIwhenSNT-Tg-1A
~ ws published.
1956 to 1963 - Nondestructive Test Operator, Babcock & Wilcox Co., Barberton, Ohio Perforned ultresonic examinations on fossil and Navy nuclear compone'nts.
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ECUCATION:
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s Graduate Troy High Scteol, Troy, W. Va..' 1954 Metall.irgy, University of Akron,1963 Cone semester)
ASMC B&PV Code and MDT (11T. RT, MT PT) Course work at Boulder, Colorado, 1963, 1968; Cleveland, Ohio,1966; Atlanta, Georgia.1970; Mt. Vernon.
Indiena 1971. 1973, 1974 Training Classes:
Educational Semincr on NDT - March 1963' SNT Seminar of NDT - May 1965 Automation Industries Seminar on Delta Technique Inspection February 1968 s
liodak Industrial Radiogrtpl1y, two week course, Atlanta, Georsria - 1970
$ Successfully completed a one-week course given by Krautkramer on the use 'of lieman nondestruct.tve extmination. techniques - 1974 Attended a one week course given by TUV (Technisher Ubesvachungs Versin) on familiarintion and application of nondestructive excminations for nuclear components fabricated for Gemany - 1973 Sucr.esdfully ccmpleted the PWR Fundementals (Construction) Course for Reactor Inspectors - Bethesda, Maryland - Noveciber 1975 Successfully completed the BWR Fundamentals (Construction) Course for RenctoY Inspectors - Bethesda, Maryland - May 1977 Successfully ccmpleted the Welding Technology and Code Course for Reactor Inspectors -
l Ohio State University, Columbus, Ohio - January 1977 Successfully ccmpleted the " Qualifying Inspectors Examination" given to Reactor Inspectcrs - King of Prussia, Pennsylvania - March 1976 Successfully ccmpleted the "Nuc' lear Power Plant Components ASME B&PV Code Section III",_
Course given by the Philedelphia Section of ASME - Philadelphia, Pennsylvania -
March 1975
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Succes:: fully completed the " Writing Skills for Scientists and Engineers" course given by the k.erican Institute for Professional Education - Novenber 1977
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Successfully completed the Concrete school at Portland Cement, Skokie,1111nois.
Succes:sf ully completed the liondestructive Examination Course given by General Dynamics in San Diego, California.
Tuct.essfully completed ti.e Electrical and Instrumentation Course given by General Electric in Chattanooga, Tennessee.
Other Sicnificant Events:
Deteloped end patented the " Electronic' Transfer" for application in ultrasonic calibration technique.
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CHARLES 5. HINSON Professional Qualifications My name is Charles S. Hinson,. I am a Health Physicist with the Radiological Assessment Branch in the Office of Nuclear Reactor Regulation of the U.S. Nuclear Eeculatory Commissior. (NRC). ! am responsible for technica1' review and evaluation of the radiation protection pr~pgram for proposed nuclear facilities. I have been with the Radiological Assessment Branch for about 61/2 years.
1 received a B.S. in Nuclear Engineering from the University of Virginia in 1974, and a M.E. in Health Physics /Huclear Engineering from the University of Virginia in 1976.
I worked in the Radiation Protection Section (RPS) of the Radiological Assessment Branch (RAB) from 1974 to 1975 and have been with the RAB since I received my M.E. degree in mid-1976. My principle function is the review of power reactor applications, both at the construction permit and operating license stage. The objective of this review is to assure that the plant is designed and operated in a canner that wil's maintain occupational radiation exposures as low as is reasonably achievable. My secondary duties include reviewing and development of HRC regulations and safety guides, participation in' task forces to resolve generic issues, and compilation of annual radiation exposure data for LWRs.
I co-authored the report entitled " Occupational Radiation Exposure at LWRs 1969-1974" (NUREG-75/032) and presented a paper entitled " Occupational Exposure and ALARA" at a AHS conference on Decontamination and Decommissioning of Nuclear Facilities.
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