IR 05000336/1992026

Insp Rept 50-336/92-26 on 920830-1106.Violations Noted.Major Areas Inspected:Steam Generator Replacement

ML20125D788
Person / Time
Site:	Millstone
Issue date:	12/08/1992
From:	Durr J, James Trapp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20125D766	List: IR 05000336/1992026 ML20125D762 ML20125D775
References
50-336-92-26, NUDOCS 9212160031
Download: ML20125D788 (12)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

STEAM GENERATOR REPLACEMENT INSPECTION REPORT / DOCKET N /92-26 LICENSE N DPR-65 LICENSEE: Northeast Nuclear Energy Company P.O. Box 270 Hartford, Connecticut 0614-0270 FACILITY NAME: Millstone - Unit 2 INSPECTION AT: Waterford, Connecticut INSPECTORS: S. Chaudhary, Sr. Reactor Engineer, DRS H. Kaplan, Sr. Reactor Engineer, DRS J, Carrasco, Reactor Engineer, DRS P. Peterson, NDE Technician, DRS N. Economos, Region Il G, Hornseth, Materials Engineer, NRR INSPECTOR: i MWQ ll > 2 5 'l L J. Trapp Team Leader',' Engineering

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Date Branch, Division of Reactor Safety APPROVED BY: #> .

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/ acqt#P. Durr, Chief, Engineerlng Branch, Division of Reactor Safety 9212160031 921204'

PDR ADOCK 05000336

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' INTRODUCTION The Millstone Unit 2 steam generators were replaced during ths < ycle 11 - 1992 refueling outage. The original steam generators were designed and fabricated by Combustion Engineering; while the replacement steam generators were designed and fabricated by Babcock and Wilcox. The replacement steam generators are similar to the original steam generators and are being installed without prior Nuclear Regu!atory Commission (NRC)

approval in accordance with 10 CFR 50.59. The replacement component or lower assembly includes the channel heads, tube sheet with the tube bundle and the secondary shell up to the midpoint on t..c transition cone. A girth weld cut was made to. separate the steam drum from the lower assembly. The steam drums were inverted and refurbished with new separators and a new feed rin Since the last engineering inspection (NRC Inspection Report 50-336/92-17), the pipe alignment / restraints were installed. As part of the setup and staging operations, both replacement steam generator lower assemblies were moved adjacent to the containment enclosure. The temporary steam drum stands were moved into the containment and assembled where the inverted steam drums would be placed for refurbishment. Cutting machines were setup on each steam generator (S/G) at the transition cone area of the steam drum. Both S/Gs were cut at the same time to separate the steam drum (S/G top head) from the lower assembly and by mid-July, both steam drums were ready for removal from the lower assemblies. Weld end preps were machined on the reactor coolant system (RCS) hot and cold legs. The new steam generators subassemblies were brought into the containment and set on the sliding bases support. The licensee verified the proper location to achieve fit-up of all terminal points. Bolts were installed on the sliding base support and torqued to the designated requirement The licensee has attained joint fit-up for all reactor coolant weld joints without excessive gaps or other dimensional deficiencies. Preheat was applied and welding of the reactor coolant system piping was completed using a narrow groove weld design. The present status of the RCS piping on steam generator No. 2 is final welding, post weld heat treatment and final radiographic testing are completed. Welding and fm' al radiographic testing are completed on Steam Generator No.1. Post weld heat treatment is expected to be completed during the week ending October 24,199 .0 BACKGROUND The principal objectives of this inspection are to assure that the steam generators are replaced in a safe controlled manner, and to assure that the integrity of the systems affected by the design change is not compromised. The focus of the inspection to satisfy these objectives is as follows:

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1 The findings of four separate inspections are contained in this inspection report. The first -

inspection _was conducted by two region based inspectors and one NRC Region 11 inspector between August 30 to September 4,1992. This inspection focused on a review of the stress -

analyses performed for the reactor coolant system cold leg pipe The second inspection was conducted by one region based inspector during September _29 to October 1,1992. This inspection focused on the welding of the reactor coolant system ]

pipm A third inspection was conducted by two region based inspectors on October 19 to October 23,1992. This inspection reviewed the post weld heat treatment records and procedures for the reactor coolant system hot and cold leg pipe welding. : A follow-up inspection was conducted of the RCS pipe stress issue. This inspection also assessed the welding aspects of the steam generator girth welds, h

A fourth inspection was conducted by one region based inspector of the nondestructive examination tests and was conducted on November 4-6,199 .0 DETAILED INSPECTION FINDINGS Reactor Coolant System (RCS) Stress Analyses (inspection Procedure IP 3_7700)

As part of the Steam Generator Replacement Project (SGRP) at Millstone 2, the cutting activities of the reactor coolant system piping from the steam generator nozzle began in the second week of July. After the cutting of the reactor coolant system lines, unpredicted pipe movement was experienced on the cold leg piping. The cold legs moved in an upward and_

outward direction from the S/G nozzle. This unpredicted movement was in the order of three quarters of an inch. The purpose of this inspection'was to evaluate the extent to which the licensee has established that stresses in the RCS piping after welding will be acceptable for service. To accomplish this purpose, the inspectors focused on the cutting of the pipe, the reported pipe movement, related stress analyses and engineering evaluation Prior to cutting, the cold leg piping structural model was a beam fixed at both ends. After cutting, it became a cantilever off the reactor vessel. The licensee had provided a vertical restraint for the reactor coolant line. Therefore, the movement of the cantilever was limited to the x and z plane orthogonal directions. The results of the licensee contractor's computer

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analysis suggested that the observed displacement toward the steam generator _ was-

- accompanied by a load redistribution. The physical observation indi_cated that the some

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displacement along the pipe was blocked, which limited the free displacement of the

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cantilever, preventing a complete load redistribution. Thus, the compressive load induced in the cold leg pipe and the steam generator nozzle parallel to' the pip.: was reduced, but not eliminated. The hot leg was also evaluated by the licensee. But, because of the geometry and the stress levels, greater analytical emphasis was placed on the cold le ,

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To assess the effect of this residual stress in compression, the licensee stated that they would perform the root cause analysis to determine the cause of the pipe movement and perform analyses to determine what force it would take to align the pipe with the center line of the S/G nozzle for welding fit-u ,

The inspectors reviewed the ABB/CE Design Report No. N-MECH-DR-002. This report analyzes stresses of the cold leg piping for the additional loads induced by rep!acemer' .; team generator fit-up and the revised seismic loading. The purpose of this analysis was to demonstrate that the calculated stresses were within the limits of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section Ill for class 1 components. The loads included the internal pressure and dead weight, the replacement steam generator fit-up loads, and the revised seismic loads. The elastic method and the ,

simplified clastic-plastic method of analysis were used. For conservatism, the maximum loading combination and stress enveloping approach were applied. . Data was used from the -

Stress Report No. CENC-1192, " Analytical Report For Northeast Utilities Service Company Millstone Point Station Unit No. 2 Piping," dated September 1973, to determine the critical stress locations in the piping syste The inspectors found these methods of analyses of the design bases acceptable for an ideal as-constructed system with zero stresses. But the zero stress analysis did not reflect the as-found configuration. The licensee was requested to justify the use of above calculation in the analysis and evaluation of the RCS loops piping for the existing state. The original and-subsequent stress analysis had assumed the pipe to be free of any construction residual stres However, the severance of hot and cold leg RCS piping indicated the existence of substantial stresses in the piping. After observing this unexpected stress condition, the licensee had neither taken steps to determine the origin of the unanticipated stress nor had any program / schedule to determine the root cause of this problem before the attachment of the piping to the new S/Gs. The approved installation procedure and schedule indicated that the piping would be moved to the original or S/G new nozzle fit-up position by hydraulic jacks and welded to the new S/G nozzles. Again the calculation analyzing this forced movement of piping for proper fit-up only evaluated the stresses induced by the hydraulic force; it did not account for the preexisting stresses that remained in the piping due to temporary restraints installed to prevent and limit the movement of the pipe after severance from S/G nozzle In response to the inspectors questions, the licensee acknowledged the need for further analysis, and committed to further review and evaluate the as-found condition of the piping before welding the piping to the new S/G Based on the review of the analysis, evaluations and discussions with the SGRP personnel, the inspectors concluded that the licensee had not performed thorough evaluation of the RCS piping movement, and the analysis and engineering calculations performed by the licensee to assure conformance to the original design basis were technically deficien _ _ _

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However, the licensee has indice.ted that additional analyses and evaluations would be performed to determine the root cause of the unanticipated pipe movernent and establish that the as-built corfiguration of the RCS piping system was in accordar.cc with the design basis and applicable piping codes. The licensee provided the NRC infor' nation regarding this issue in a letter to the Region I Regional Administrator dated Septembu 23,1992. Attachment 3 of this letter outlines a six step plan to address this issu The actions taken to address this issue were followed up during an inspection conducted on October 19-23, 1992. The inspector interviewed the Supervisor of the Corporate (Berlin office) Pipe Stress Group and determined that the licensee has investigated the root cause of the observed pipe movement. The major contributors for the pipe movement were Jesign dead weight of the piping, construction weld shrinkage and out of sequence stress relief of the cold leg piping during original construction. To address the displaced pipe, the licensee has restored the reactor coolant lines to their unrestrained state and measured the maximum displacements. The licensee has indicated that the net displacement has been input as " cold spring" in the piping stress model. Prior to the fit-up and welding of the RCS, the licensee obtained preliminary results of their analysis that indicated that the stresses were within code allowable and nozzle loads were acceptable. The inspector determined that this approach was acceptable and is consistent with the licensee plan that was submitted to the NRC in a letter dated September 23,199 Based on the overall progress of the project and the interviews with engineering personnel regarding the pipe stress of the RCS piping, the inspector concluded that the licensee is properly controlling welding of the Unit 2 steam generatar's RCS piping and they are taking the proper steps to assess, analyze and correct the reported movement of the RCS piping such that the final RCS pipe stresses will be reaffirmed to be within ASME Code limit The acceptability of the RCS's structural integrity stress analysis remains unresolved pending the review of the final analyses (NRC Unresolved Item 50-336/92-26-01). Welding Inspection by headquarters and regional inspectors of the S/G cold and hot leg nozzles to the existing RCS piping and the new lower assembly to the existing steam drum was performed on September 29,1992 to October 1,1992 and October 13-14, 1992. The following observations were derived from these inspection .2.1 RCS Welding The purpose of this inspection was to examine the field application of narrow gap tungsten inert gas (TIG) welding. This technique is being applied to the circumferential welds rejoining the existing RCS pipe to the new steam generator nozzles (hot and cold legs). This inspection is a follow-up of a previous site visit during June / July 199 _ _ _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _

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During the initial visit, the pre-outage activities (such as mock-ups and welder qualification program) and the necessary code requirements (such as welding procedures) were reviewe Because the narrow gap technique is relatively new in Held applications at power plants, assessment of both the welding technique and the licensees' management of the welding process were conducted. Based on the initial inspection, it appeared that the licensee has an effective welding program that would enable successful application of the technique. This was due to the experienced welding engineer who had been dedicated to the project, and that a thorough series of mock-ups had been set up and used to prove and refine techniques and train the welding operators. Additionally, the narrow gap technique appeared to be an effective method to rejoin heavy wall pipes, especially when weld distortion and/or weld

- 3hrinkage are major issue Welding of the first steam generator hot leg was in progress on September 29,1992. The start of welding was substantially delayed by the unexpected discovery of designed-in cold pull in the RCS pipes. This caused greater than expected difficulty in obtaining the proper fit-up prior to making the open butt gas tungsten arc welding (GTAW) toot pass. Accurate fit-up is mandatory to the success of the narrow gap method due to the close tolerances required of the groov The narrow gap joint base materials are P-No. I carbon steel. The pipe is SA 516, grade 70 plate rolled and welded with stainless cladding. The S/G nozzle is forged SA 508, class 1 with stainless cladding. GTAW filler wire is ER70S-2. Cladding in the weld end prep area will be restored by the automatic GTAW process sometime after the girth weld has been completed to the point that the first informational radiographs show satisfactory results. This was not yet underway during this inspection due to the need for some unanticipated inside diameter weld build-up with carbon steel to compensate for thickness mismatch resulting from heavier than expected wall thickness of the new nozzles. This build-up was being performed with the same automatic GTAW equipment that will be used for the stainless claddin The groove had been filled about one inch (out of approximately 3% inches total wall thicknest when welding was stopped for an in-progress radiograph. Preheat (200 *F minimum, was maintained during radiography. A panoramic view with the source (Iridium 192) inside the pipe was made along with single wall views with the film outside the pipe. These informatmnal (not record) views are intended to catch unacceptable flaws before they become too deeply buried for expeditious repair. All Hims and reader sheets were reviewed by the staff. No unacceptable indications were detected during this revie Production welding was monitored with two television cameras per welding head. One camera shows the weld puddle in the direction of travel along with the wire feed into the molten puddle. The other camera shows the trailing edge of the puddle. With the two views, wire feed, arc stability, oscillation, puddle size, the sidewalls, and the solidified bead were all clearly visibl ____ _ - _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ .

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Since weld progression is uphill (roughly 6G position), two machines (Diametrics) are in simultaneous operation per joint. This enabled the observation of a number of starts and stops. Establishment of the arc and puddle and initiation of wire feed were well controlled and clearly visible. Problems with wire feed or oscillations were readily observable and immediately corrected. Tie-ins to the previous bead or the opposing bead at the stop position (in the 12:00 position) were well controlled. No problematic cratering occurred at the stops when the arc was broken. Any coarse ripple or excessive deposit was readily removed by local grindin One wire feed failure was observed. The operators were readily able to observe the feed failure and perform a controlled stop. Festart was simply a repeat of a normal start after the torch technician cleared the problem. 'i a inspectors concluded that the welding operations were well controlle After another 1% inches depth of deposit, welding was stopped for more informational radiographs. All fihas and reader sheets were reviewed by the inspectors. Two acceptable indications of porosity were detecte Because of the large forces required to force the pipe into proper fit-up, questions about restraint during the post weld heat treatment (PWHT) operations were discussed with the licensee's stress analysis group. No cold pull restraints are intended to be used during the PWHT. This is due to the fact that the amount of cold pull required (by calculation) results in a stress of roughly 400 psi at the weld joints. The amount of stress in excess of this value that is due to the fit-up operations is still within code allowable. Any fit-up stress that is lost i

during PWHT due to creep relaxation of the pipe while it is at 1100 degrees will be intentiona One problem that resulted from having the cold spring was the difficulty in jacking the pipe into proper fit-up. This is a difficult task when the numerous interferences inside con ainment are encountered. The licensee experienced difficulty in locating suitable anchor points for the jacking equipment. Despite these problems, fit-up of one cold leg was l accomplished during the inspection by means of jackin The inspector reviewed the authorized nuclear inspector's (ANI) hold points for fabrication of the RCS and the SG steam drum to the new lower assembly girth welds. The points covered essential activities such as fit-up, magnetic particle testing (MT) and liquid penetrant testing l

(PT). The licensee provided the inspectors with the current log sheet for welding the hot leg of SG-2. The recorded welding parameters conformed to the welding procedure specification (WPS) requirements.

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The inspector concluded on the basis this inspection as of October 2,1992, that the RCS pipe and head to shell welds were being fabricated in accordance with ASME Class NB requirements under the close supervision of Northeast Utilities Service Company (NUSCO)

f welding, nondestructive examination (NDE), and quality assurance (QA) personnel.

j 3.2.2 Post Weld Heat Treatment The inspector reviewed the preheat and final post weld heat treatment charts for the hot leg of SG-1. Heat treatment was performed by Cooperheat in accordance with Procedure 31690-CHP-027, Rev. 5. The chart showed a continuous preheat cycle of 200 F, and a PWHT of 1150*F for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 45 minutes with no significant variations. The heating and cooling rates as determined from the chart did not exceed the maximum specifed rate of 100 F in any hourly interval. The PWHT cycle was monitored by 12 thermocouples located on the outside diameter of the joint. A calibration certification validating the accuracy of the heat treat recorder, dated July 7,1992, was submitted to the inspector for review. The test temperature ranged from 202 F - 1602 F. The calibration standards were traceable to National Institute of Standard .3 S/G Girth Welds The inspector witnessed the in-process welding of the girth welds joining the SA-533 grade B, low alloy steel steam drum to the new lower assembly of S/G's 1&? using the flux core process in accordance with WPS 3-3-F2H. The root pass and a few succeeding layers were first deposited using the manual gas tungsten are process, followed by the shielded metal are process in accordance with WPS-3-3-BA1. Visual examination of the partially completed welds a;,peared to show a good fusion with no obvious defects. The initial in-prxess radiography of SG-2 revealed tw.- linear type indications. The area containing these indications was scheduled for light grinding. Radiography was to be performed prior to and after final PWH .4 Ultrasonle Inspection An independent ultrasonic examination was performed by the NRC of weld number P-11-C-1-B, on the cold leg piping of the reactor coolant pressure boundary and weld number P-10-C-3-B on the hot leg of the reactor coolant pressure boundary of steam generator number two. This examination was performed on the welds in their final condition utilizing the licensee's procedure for ultrasonics identified as: NU-UT-26, " Ultrasonic Examination Primary Coolant Pipe Welds - Millstone Unit 2;" dated 10/15/92 with Change Notice Number (CNN): NU-UT-26-1, dated 10/23/92 and CNN: NU-UT-26-2, dad 10/23/9 The NRC inspector utilized the licensees calibration block UT-60 and UT-15 for the purpose of calibrating the NRC's ultrasonic instrument. The equipment, used by the NRC, closely matched the licensee's cauipment in performance and calibration characteristic RI

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The ultrasonic examination performed on these welds was for die pogose of satisfying the reouirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section IX,1980 Edition with Winter 1980 Addenda for preservice inspectio The licensee's final report for weld P-10-C-3-B (RCS-SG#2 hot leg), called an " ultrasonic indication data sheet" (M2-92-159-UT), listed 54 indications: 9 as " spot indications = no length observed" and 45 as reflectors without any characterization. The report did not correlate the spot indications to the porosity indications that were revealed by radiograph The inspector was not able to determine, from the licensee's ultrasonic test report, which indications were porosity, thus making the direct comparison of the NRC's independent results with those of the licensee's very difficult. This was of concern to the NRC inspector since it is the goal of the preservice inspection to characterize original indications for direct comparison and analysis with the results of future examinations. If it were difficult to correlate the NRC's independent evaluation to the current ultrasonic report the possibilities of correlation in the future would diminish with time and the unavailability of the original ultrasonic technicians. This concern was brought to the attention of the licensee's representative and a memorandum was issued by the licensee's Ixvel III inspector in ultrasonic testing, with suf6cient detail to fully portray the state of the indications in this weld. In all other regards, the NRC's ultrasonic evaluation of the welds compared favorably with the results obtained by the license .5 Radiographic Inspection The radiographs of weld number P-10-C-3-B were evaluated in order to arbitrate the results of the ultrasonic inspection. It was noted by the inspector that a linear indication contained in one of the radiographs was not listed on the radiographic evaluation report (also referred to as a reader sheet). When this was brought to the attention of the licensee it was determined that the indication had not been evaluated for disposition. Every indication must be listed on the reader sheet in order to assure that an indication has been evaluated. Failure to list and dispose of an indication on a report should be interpreted as a failure to evaluate the indication. This requirement was emphasized in this case because the licensee determined that the indication had not been evaluated until the NRC identified it. Appendix IX of the ASME Boiler and Pressure Vessel Code,Section III,1971 Edition with addenda through summer of 1971, for nondestructive examination methods, under paragraph IX-3340 requires that "the Manufacturer shall record on a review form the interpretation and disposition of each film." In conformance with this requirement the licensees procedure for radiography MP-XII-08: " Radiographic Examination of Weldments and Materials," Revision 0, dated 8/14/91, with procedure change notices 001, dated 5/30/92; 002, dated 9/26/92; and 003, dated 10/27/92; states in Exhibit A, Revision 0, " Instructions for Preparation of Radiographic Reports," Line 15: that acceptable indications that were evaluated or comments which concern interpretation shall be recorded under the remarks section of the report. The failure to record this indication is a violation of NRC requirements (NRC Violation 50-336/92-26-02).

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) It was also noted that a light density image of a lead letter "B" appeared on one of the radiographs. The presence of 6is lead letter "B" is an indication of excessive backscatter

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radiation. The presence < excessive backscatter radiation can be deleterious to the sensitivity of the radiograph. The licensee's document: " Millstone Unit No. 2 Steam Generator Replacement;" Plant Change Number 2-036-91, Rev. O, Safety Evaluation, Section 3D, Part 5 states that radiography will be in accordance with the ASME Boiler and Pressure Vessel Code, Section Ill,1971 Edition with addenda through summer of 1971. Appenc'ix IX of this issue of this section of the code, for nondestructive examination methods, under paragraph IX-3324 requires that " objectionable scatter radiation shall be reduced by suitable filtration."

In conformance with this requirement the licensees procedure for radiography MP-XII-08:

" Radiographic Examination of Weldments and Materials," Revision 0, dated 8/14/91, with procedure change notices 001, dated 5/30/92; 002, dated 9/26/92; and 003, dated 10/2'/92; states: "If a light density image of the lead letter B appears on the film, backscatter is then evident and a new exposure must be made with adequate shielding." The licensee's procedure in Paragraph 4.0 of Appendix 2 states: "If a light density image of the lead letter B appears on the film, backscatter is then evident and a new exposure must be made with adequate shielding " The presence of the light image of a lead letter B on the radiograph is a violation of these requirements (NRC Violation 50-336/92-26-03). This was brought to the attention of the licensee. The licensee had corrected other instances in the same radiographic sequence where the lead letter B indicated excessive backscatter radiation and committed to making the conection in this case as well. In all other regards the radiographs were in compliance with the requirement .6 Quality Assurance The inspector reviewed filler material certifications for the RCS piping welds and the head to shell girth welds as provided by Fluor. The certification for the RCS filler material indicated that the material (heat No.2327F-ER70S-2) as furnisheu by U.S. Welding Corp. exhibited satisfactory tensile and yield arength as well as +10'F charpy test results after 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> of simulated PWHT. The certification for the shell-to-head girth welds indicated that the material (E81T1-N1) (heat 31162) as furnished by alloy rods exhibited satisfactory tensile, yield strength, and charpy impact values at -20 F after 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> simulated 1125 F PWH A typical field requisition slip for the subject points was checked and was found to match the filler material certificatio The inspector reviewed typical NDE liquid penetrant (LP) and magnetic particle (MT) reports for the RCS piping. No discrepancies were found except it was noted that incomplete fusion on the root side of SG-1 hot leg was found as noted in report 92 MT-153 on 10/10/92, subsequently cleared by grinding and reinspected. Report 92-MT-154 indicated that the area was considered acceptable. Also noted were MT reports after removal of spot welds attaching welding machine tracks to pipes and removal of thermocouples in addition to final MT _

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in The inspector reviewed a sample of 682 surveillance reports provided by the NUSCO Quality Services Department that covered all activities in the S/G replacement program, including fit-l up, welding and NDi!. Although, several problern areas were four.d early on the project involving welder training and electrode control as managed by 1:luor, no significant problems were reported with the 1(CS piping and girth welds. The surveillance; r ports for these welds were detaiH and provided wide cove. age of essential activities includind renonnel certi6 cation .0 M ANAGEMENT MEETINGS The inspectors met with those denoted in A/achment 1, on September 4,1992, October 23,1992 t: d November 6,1992 to discus, the preliminary inspection Ondings that are detailed in this report.

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A'ITACIIMENT 1 tersons Contacled NORTilEAST !1TILIIJES R. lilanchard, Millstone 11 F. Libby, Quality Control Services Sulwrvisor T. Manson, SAE G. McElhone, QSD SGRP R. Nccci, Project Manager S. Orefice, Project Engineer F. Perdome, EBASCO J. Resetar, Enginecting Supervisor J. Rhodes, Senior Mechanical Engineer A. Silvia, Senior Welding Engineer R. Thomas, SGRP ELUOR DANIELS b. liarper, Senior Welding Engineer U.S. NUCLEAR REGULATC AY COh1MJSSION D. Dempsey, Resident inspector

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