ML19275A815
| ML19275A815 | |
| Person / Time | |
|---|---|
| Issue date: | 09/28/1979 |
| From: | Donna Anderson, Phillip G, Potapovs U NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML19275A812 | List: |
| References | |
| REF-QA-99900529 99900529-79-1, NUDOCS 7910190143 | |
| Download: ML19275A815 (49) | |
Text
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U. S. NUCLEAR C0?e!ISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION IV Report No. 99900529/79-01 Company:
Southwest Research Institute San Antonio, Texas Dates of Investigation: August 9, 20, 21 and 22,1979 Investigation at: San Antonio, Texas Investigators:
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h'2k'79 D. G. Anderson, ?rincil al Ihspector Date Program Evaluation Sec ion Vendor Inspection Branch Y
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9-28 19 G. A. Phillip
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Date Investigation SpecialistI i
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Uldis Potapovs, Chief Date Vendor Inspection Branch Investigation Summary:
Investigation on August 9, 20, 21, and 22, 1979 (Report No. 999000529/79-01)
Areas Investigated: Because allegations were made that: (1) a certification record had been falsified; (2) an ultrasonic examination had been improperly performed; (3) a technician did not have required education; (4) another firm had not recorded indications during baseline nondestructive examinations which were subsequently identified during examinations performed by SRI; (5) instruments used on a seismic qualification test for NRC were not calibrated; (6) contaminated items were shipped to SRI from nuclear power plants: and (7) nothing was done about a nozzle crack identified during an inservice inspection, this investigation was performed to review procedures and records and conduct interviews of personnel.
The investigation involved thirty-one investigation hours by two investigators.
Results: None of the allegations were substantiated. No items of noncompliance were identified.
1180 090 190lN 3 7910
REASON FOR INVESTIGATION On August 6 and 8, 1979, an individual visited the Region IV office and made allegations relating primarily to nondestructive examinations performed by Southwest Research Institute (SRI) at nuclear reactor power plants.
SUMMARY
OF FACTS During visits to the Region IV offices on August 6 and 8, 1979, and during a subsequent interview on August 9,1979, an individual mad? seven allegations primarily regarding the nondestructive examination activities of SRI which performs such activities at several nuclear power plants and as a contractor to the NRC. An investigation of these allegations was conducted at SRI on August 20-22, 1979. None of the allegations were substantiated.and no items of noncompliance were identified.
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DETAILS 1.
Personc contacted
- Martin Goland, President A. W. Betts, Senior Vice-President, Operations
- G. N. Van Steenberg, General Counsel
- C. E. Lautzenheiser, Vice-President, Quality Assurance Systems and Engineering Division
- A. R. Whiting, Executive Director, QA Systems and Engineering Division
- W. T. Flach, Director, Department of Engineering Services U. S. Lindholm, Director, Department of Material Sciences L. S. Albrecht, Manager, Quality Ass,urance
- D. F. Rosow, Manager, Nuclear Field Services Section D. G. Cadena, Jr., Radiation Safety Officer R. E. Engelhardt, Senior Research Engineer C. L. Cotton, Inspection Engineer W. T. Clayton, Technical Activity Supervisor J. G. Godwin, Supervisor, Technical Liasion D. C. Schtidt, Engineering Technologist D. Cantello, Senior Technician H. L. Clark, Senior Technician R. L. Spinks, Senior Technician L. J. Kasper, Senior Technician W. Heinemeier, Senior Technician B. Richter, Secretary M. Risner, Secretary 2.
Introduction On August 5,1979, articles appeared in two San Antonio, Texas, newspapers concerning allegations made against the Southwest Research Institute by a former employee. On August 6,1979, the alleger ;rought the allegations to the NRC Region IV office in the company of two Dallas newspaper reporters. During that visit the alleger agreed to provide a written statement containing the details of these allegations. On August 8,1979, the written statement was provided by
' alleger, a copy of which, with names deleted, is attached to this repet. as Exhibit A.
To obtain additional details an interview with the alleger was conducted on August 9, 1979, at his residence in San Antonio, Texas.
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5 3.
Allegations Based on the information provided during contacts with the alleaer on August 6, 8, and 9,1979, seven allegations were derived.
These allegations and the information obtained regarding them during the investigation are set forth below.
Allegation 1 - A certification of magnetic particle material used at the Brunswick Steam Electric Plant in 1974/75 was falsified.
The alleger stated that as an SRI employee in late 1974 or in early 1975 he had participated in an internal review of all documentation relating to work performed by SRI at the Brunswick facility.
This review was intended to determine whether there were any required documents which were missing or incomplete. During this review he could not find the batch number for the magnetic particle material used in the magnetic particle testing performed at Brunswick.
In this connection he could not find a manufacturer's certification that the material contained less than 1% halogens and sulphur.
After searching unsuccessfully for three days he was instructed by one or both of two named individuals to prepare a certificate which would state that the material contained less than.01% halogens and less than.01% sulphur. He was instructed to identify the materials supplier as Magnaflux Corp. and to backdate the document.
He prepared the document as instructed and had a secretary type it up for him.
Since the document required a second signature he explained the situation to another technician and he attempted to get him to sign it also. He gave the falsified document to the technician.
He said that some time later he saw the document in a file drawer, the location of which he described in detail.
When he saw the documect in the file drawer he noted that only his signature was on it.
Finding. On August 20, 1979, both individuals whom the alleger identified as having instructed him to prepare the false document certification were interviewed. Both denied ever having requested or instructed anyone to prepare any false documentation of any kind.
SRI personnel stated that magnetic particle testing was performed only on the reactor vessel closure studs at Brunswick.
The documentation relaticg to this work was reviewed.
It was noted that the work was performed on successive days during the period July 31 through August 8,1974, except for August 4,1974, which was a Sanday.
The data sheets completed in the field show the number 20A as the batch number.
A copy of the manufacturer's mixing instructions identifies No. 20A as Water Dispersable Magnaglo. On the basis of the site generated documents, Magnetic Particle Examination Data Sheets were prepared and both sets of documents are filed together.
It was noted that the data sheet identify the material batch number as 2LO64.
A reproduced copy of a certification of materials furnished i180 093
6 by Magnaflux Corporation for this material was provided by SRI and is attached to this report as Exhibit B.
It was noted that no date appears on this certification. Since it is possible the original document bore a date stamp that did not reproduce, the original was requested. SRI personnel stated that they had been unable to locate the original. They also advised that they had contacted the Magnaflux Corporation and had been advised that they could not locate the certificatiot in their files either.
Each of the above-mentioned data sheets, serially numbered 18001-18011, bore the following handwritten notations with variations in the dates: " Note: This sheet filled out'by (alleger's name) based on field data.
(A11eger's initials)."
"The above date (29 Jan. 75) represents the date that this sheet was filled out for the examinations performed on August 2,1974.
(Aliager's initials)."
Included in the data package for the Brunswick work was a memorandum dated February 14, 1975, signed by the alleger stating that the mixture used at Brunswick was No. 20A, Water Dispersable Magnaglo; that the mixture included one part Magnaglo No. 14A, Batch No. 2LO64; and that mixing was done according to the supplier's instructions.
A copy of this memorandum with the alleger's identification deleted is attached to this report as Exhibit C.
According to SRI's travel records the alieger was at the Brunswick site on two occasions in 1974; June 10 to July 2 and August 5 to August 16. As indicated above, the magnetic particle examinations were performed during the period of July 31-August 11, 1974.
In all likelihood some mixing of material was done at the site just prior to and possibly during this period.
It appears therefore that the alleger's memorar.dum could not be based entirely on his personally performing or witnessing all of the mixing that was done. For this reason he may have regarded this memorandum as a false document.
This document, however, does not fit the description of the document the alleger stated he had falsified. SRI personnel were unable to state whether anyone had requested the alleger to prepare the memorandum or whether he had generated it on his own initiative.
It was stated, however, that there was no requirement for such documentation.
It was pointed out by SRI personnel that magnetic particle examinations are normally only erformed on carbon steel and that the presence of sulphur and halogtns in the material is therefore not a matter of concern.
Specifically, the closure studs at Brunswick are of carbon steel.
It was further stated that the ASME code applicable to magnetic particle testing does not require any manufacturer's certification I
1180 094
a 7
of the material used. It was ascertained that the ASME code does contain requirements regarding halogens and sulphur in the liquid penetrant material used in penetrant testing.
The Brunswick docu-mentation package contained three certifications provided by the supplier of the Magnaflux Corporation, for these materials. They were dated August 22, 1973, October 31, 1973, and April 18, 1974, respectively.
It was noted that the ASME code requirement for penetrant testing immediately followed those related to magnetic particle testing.
The secretary who was identified as having typed the document the alleger falsified and who had also typed the above-mentioned memorandum was interviewed. She stated that she could not recall typing such a document nor did she specifically recall having typed the memorandum.
This recretary, who has since been married and now occupies another position, and a secretary who now has the responsibility for the files described by the alleger as containing the falsified document, provided assistance in an effort to locate the document.
It was ascertained the file cabinet described by the alleger and other dSsociat9d file cabinets had previously been located in Room 3 of Luilding 82 but had been moved to Building 88, Room 1.
In the course of the move the file as well as their contents had been rearranged. Tbe physical location described by the alleger was therefore no longer applicable. The contents of the drawer containing documents relating to the Brunswick project were reviewed. No additional documents signed by the alleger were found.
The individual identified by the alleger as the technician to whom he had given the falsified document was interviewed. He stated that he had never been requested to sign the falsified document and that a conversation such as described by the alleger had not taken place.
He said he did recall that on one occasion the alleger had come to him with a Brunswick site generated document that the originator had neglected to sign. The alleger had asked him if it was his document and, if so, would he sign it.
The technician said he informed the alleger that he had not done the work to which the document related and he could not therefore sign it.
The technician said that he assumed the alleger did eventually find the right technician to sign the document since he heard no more about it.
It was indicated to the technician that the alleger had advised the NRC during a telephone conversation on August 17, 1979, that the technician would supply the falsified document to NRC during the investigation. The technician stated that he had been in contact I180 095
8 recently with the alleger and the alleger had discussed his allegations with him. The' technician stated he had no documents to give to the NRC and he could provide no additional information about the matter.
In sn= mary, no evidence or information w;s obtained that any re-quired documentation had been falsified.
Allegation No. 2.
During ultrasonic testing of the reactor pressure vessel head-to-flange weld at Big Rock Point in early 1975 the Level II technician had not been continueusly present and a significant indication may have been ignored.
I The alleger stated that he was a Level I (Limited) technician when he participated in an inservice inspection by SRI at the Big Rock Point facility in early 1975. Part of that activity had been the ultrasonic examination of the reactor pressure vessel head-to-flange weld. He said that a technician performed the scrubbing operation on the weld with the head located about ten feet above floor level.
A Level II technician was required to continuously monitor a scope located at floor level. Any significant indications, i.e. any re-sopase greater that 100% of the referenced level, must be recorded fur subsequent evaluation.
According to the alleger the Level II technician left the area during the examination acd the other technician asked him to watch the scope. At one point he saw a spike which may have been record-able. He said he did not know how to record it properly. When the examination was completed the technician came down from the vessel head and at that time the Level II technician returned. The alleger informed the Level II technician that he had <een a spike on the scope and indicated the examination should be redone. The Level II technician responded by saying "no indications." The alleger stated he also brought the matter to the attention of the team leader, but nothing was done and the record of the examination was completed showing no recordable indications.
Finding.
On August 21, 1979, the technician who had performed the scrubbing operation during the ultrasonic examination of the head-to-flange weld at Big Rock Point in 1975 was interviewed. He stated that the Level II technician had left the scope briefly at cne point during that examination. The technician said the alleger, who was present for on-the-job training in ultrasonic testing, left the area and had started to go down the wrong stairway. The Level II technician went after him since he was responsible for him. When the Level II technician returned to the scope the technician went back to a point on the weld he had been scrubbing before the Level II technician left the scope and then proceeded to scrub the rest of the area to be examined. He stated that he was convinced the ultrasonic examination had been properly performed and it was his recollection no recordable 11'80 096
9 indications were found. He stated that at no time during this examination did the alleger make any mention of having seen a spike on the scope.
The technician went on to say that he mentioned the occurrence to the crev leader who instructed him to keep an eye on the Level II technician to assure he watched the scope as required.
If he noted that the Level II technician was not watching the scope at any time, he was instructed to repeat that part of the examination to be sure it was done properly.
On August 21, 1979 the Level II technician who performed the head-to-flange ultrasonic examination at Big Rock Point in 1975 was interviewed.
He stated that on one occasion during that examination he had left the scope because the alleger had walked away from the work area and had started to go down into an area which was a high radiation area.
Since the alleger was assigned to him to gain field experience on ultrasonic examinations, he was responsible for him as well as the work. He therefore went after the alleger and told him to stay near the scope. When the Level II technician returned to the scope with the alleger, the weld examination was begun over again to assure the examination was complete. He said he was certain the examination was properly conducted and no recordable indications were noted.
The individual who was the crew leader during the 1975 inservice inspection at Big Rock Point was interviewed on August 21, 1979.
The crew leader said he was not in the area wb' I the head-to-flange weld was examined. He recalled, however, tha c the individual who had done the scrubbing had expressed concern.'out the Level II technician not watching the scope. He said he had instructed the individual to watch the Level II technician and if he saw him look away or leave the scope he should go over that part of the weld again.
The crew leader said he regards the Level II technician as a competent and conscientious technician. He said that he was confident that the ultrasonic examination was performed properly and stated that he had received no information that a spike had been seen during the examination. He further stated that the Big Rock Point personnel had performed an audit of their activity during this ultrasonic examination and they had said it had been performed according to procedures.
A copy of the SRI report on the 1975 inservice inspection was obtained from Big Rock Point.
Included in this report was Sheet No. 211001, fressure Vessel Weld UT Examination Record, which is a record of the head-to-flange weld examination.
This record shows there were no recordable indications and that the examination was performed on January 25, 1975, by the two technicians interviewed.
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10 A copy of an internal memorandum audit report dated February 28, 1975, prepared by a Big Rock Point employee was also obtained from that facility. The memorandum states among other things that on January 25, 1975, SRI personnel were observed inspecting the closure head flange and that no deficiencies were noted.
Allegation 3.
An individual was not properly qualified as a technician because he did not have a high school diploma or its equivalent.
The alleger stated that on his application for employment at SRI he had falsely indicated he was a high school graduate. About two weeks after being employed he discovered a high school education was a minimun requirement for certification as an inspector. He informed his supervisor of the problem and offered his resignation but it was refused. He was instructed to get the equivalent of a high school diploma, a GED, which he acquired in September 1974. During his visit to the Region IV offices he stated that he had performed work at Brunswick as a technician before he received the GED. During the August 9 interview he said he had not performed any work at nuclear power plants prior to September 1974.
Finding.
Or. august 20, 1979, SRI advised that the alleger was employed from May 20, 1974, to September 16, 1977. A copy of his GED showed it was issued by San Antonio College on September 23, 1974. SRI
-advised that available, but not necessarily complete, travel records showed the alleger had been at Brunswick from June 10, 1974 to July 2, 1974 and from August 5, 1974 to August 16, 1974.
SRI files contained the following certifications for the alleger which were dated prior to September 23, 1974.
Date Certification Level 5/24/74 Magnetic Particle Testing I Limited 5/24/74 Ultrasonic Testing 1 Limited 5/24/74 Penetrant Testing I Limited 8/29/74 Penetrant Testing 1
8/30/74 Visual Testing 1
9/3/74 Magnetic Particle Testing 1
The alleger's former supervisor confirmed that the alleger did inform him that he had falsified his application regarding his education. He talked the matter over with his supervisor, and although falsification of an caployment application is a basis for terminating an employee, it was decided that the alleger would be retained if he obtained a GED. The former supervisor stated he could not recall when the alleger came to him about the matter.
11 Regarding the alleger's certification, the supervisor stated that in 1974 ASME had no trainee level. SRI had used Level 1 (limited) as a trainee designation even though that designation was not officially recognized. This designation meant that an individual had received a technical orientation and had no pre-requisites regarding previous formal education.
The supervisor said that two visits were made to Brunswick by the alleger when he had no certification beyond Level 1 limited. He was tent to Brunswick to gain on-the-job field experience and did not conduct any independent testing, interpret any results of tests or write any test reports.
Allegation No. 4.
Baseline examination by another firm showed no indications, but SRI recorded several indications during subsequent examinations.
The alleger stated that another firm whose identity he could not recall had performed the baseline NDE work at Turkey Point 3.
The utility, Florida Power and Light Company, became dissatisfied with their work and retained SRI as the NDE contractor and requested them to check the other firm's work. When some of the SRI technicians returned from Turkey Point they told the alleger that they had found several indications where the other firm had recorded nothing. They said that the previous firm must have falsified their work.
Finding. SRI personnel advised that it was their understanding that Florida Power and Light Company had been favorably impressed by the verk SRI had performed for them at their St. Lucie plant.
It was for this reason that SRI was awarded the contract for the NDE work at Turkey Point.
It was indicated that Westinghouse had performed the NDE work at Turkey Point before SRI was engaged to do it.
According to SRI, the Westinghouse approach to the work differs from theirs but both meet the ASME requirements. Westinghouse met the ASME requirements that greater than 100% DAC are recordable indications. In performing the examinations there are some indications that require the exercise of judgment by the technician on the job as to whether a given indication is recordable for subsequent evaluation and disposition. To relieve their technicians of this burden and to permit the making of these judgments in a more conducive environment, SRI established a lower threshhold, greater than 50% DAC, for their technicians to use in recording indications. The indications recorded by SRI technicians are subsequently reviewed, evaluated and dispositioned back at the SRI facilities in San Antonio.
I180 099
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12 SRI emphasized that the difference in the number of indications recorded at the time of the tests are the result of the difference in approach but that the end result is the same and they had no information to suggest that the work by Westinghouse had been inadequate or unacceptable.
Allegation 5.
Instruments used in a seismic qualification test performed under an NRC contract were not calibrated.
The alleger stated that in 1977 he was transferred into the Department of Mechanical Sciences. While working in that department he became aware that tests were conducted using equipment that did not have calibration stickers or which had stickers which indicated expired calibration dates. He said he was not concerned about the stickers themselves but that the equipment had not been calibrated as required.
He indicated the testing program began in June 1977 and was conducted under contract with the NRC to establish a type acceptance code for cabinets containing instruments used in reactor facilities. He indicated the welds of the cabinets were being seismically tested.
He indicatzed that the tests he referred to were preliminary tests and that before subsequent tests were conducted the equipment was calibrated. He indicated he had discussed the matter with two QA-QC personnel and they had said the equipment should be calibrated even for these early tests.
Finding. On August 21, 1979, the two individuals with whom the alleger stated he had discussed the matter were interviewed. One individual stated 'ae had no recollection of any discussion with the alleger concerning the calibration of instruments. The other individual recalled that the alleger had expressed a concern about the use of uncalibrated instruments on an NRC contract activity to him. The individual stated that he looked into the matter and deteomined the contract requirements did not stipulate that the instruments used in the tests referred to by the alleger be calibrated.
He said he satisfied himself that there was no problem but he did not inform the alleger or make any record of his finding.
From a review of a report, SWRI Project 02-4675, NUREG/CR-0345, An Evaluation of Seismic Qualification Tests for Nuclear Power Plant Equipment, Final Report, September 1, 1976-August 31, 1978, which was submitted by SRI to the NRC, a list of the equipment used in the tests was obtained.
Procedure XII EE-101-0, Calibration of Mechanical Sciences Dynamic Test Equipment, was reviewed and it was determined that the procedure requires that the instrumentation be calibrated at certain intervals and that the calibration results be recorded. The procedure does not require that the calibration stickers on the instruments be kept up-to-date.
1180 i00
13 Equipment Calibration Records for the following instruments were examined:
Controller, SWRI, S/N 2 Accelerometer, B&H 4-202-001, S/N 19746 (Vertical)
Accelerometer, B&H 4-202-001, S/N 19742 (Horizontal)
Seismic Table-SWRI, S/N 1 Servo Controller, Team Corporation, Model 1522, S/N 14335 Response Spectrum Analyzer, SD 13231, S/N 12 Tape Recorder, Ampex FR-1800, S/N 7040122 These records show that all the inst,rumentation was calibrated on a timely basis during the period the tests were conducted. Another instrument used, an X-Y plotter, is not subject to calibration.
Personnel involved in the testing activity also pointed out that these tests were conducted for the purpose of determining if the methods of testing were adequate. The inspector contacted the Water Reactor Safety Branch of the USNRC to confirm that the tests were not conducted for the seismic qualification of equipment for a nuclear power plant. The Chief of this branch also indicated that the work that SWRI has performed for them has always been completely satis-factory.
Allegation 6.
Contaminated items were shipped to SRI from nuclear power plants.
i The alleger said that vessel and pipe standards which had been shipped from Turkey Point and Brunswick were received at SRI in a contaminated condition and the shipments did not bear appropriate radiation labels. He said the levels of contamination were on the order of 100 cpm (counts per minute) and one of the technicians was upset about this contamination.
Finding. On August ~21, 1979, the technician identified by the alleger was interviewed. He stated that he is a very outspoken individual and often expresses his displeasure about things that are of no great significance, He stated he had no recollection about this specific matter.
"c said that all items taken or shipped from nuclear power plants by SRI personnel are checked for contamination and cleared by site radiation protection personnel. He said he could recall no exceptions to that requirement at any site he has worked.
On August 21, 1979, the SRI Radiation Cafety Officer (RS0) was-interviewad. He stated that SRI procedures require that all items used in inservice inspections at nuclear power plants must be checked i180 101
14 for contamination upcn receipt at SRI before they are released for reuse. He said that occasionally a slightly contaminated item is received from a nuclear power plant which was released by the site as clean. He said the higher radiation background at the site location would account for the fact that no contamination is detected there while a few counts per minute are detected upon receipt at SRI.
The RSO also stated that SRI is a licensee of the State of Texas, an agreement state, and as such the State of Texas approved the SRI Radiological Health and Safety Manual. Section 8.1.4 of this manual allowsSRItoreceiveandreleaseequipmentforusewithoutdecongamina-tion, where wipe test results indica'te less than 0.005 uCi/100 cm.
The RSO indicated that 100 cpm on a survey meter is less than 0.005 uCi. The contamination levels cited by the alleger are within the limits allowed by state and federal regulations for safe shipment, handling and storage.
Allegation 7_.
During an inservice inspection at Pilgrim in 1976 a crack-like indication on a nozzle was found and nothing was done about it.
The alleger stated that during an inservice inspection at the Pilgrim nuclear power plant conducted by SRI in February 1976, a crack-like indication was found on a feedwater nozzle.
A representative of SRI management, as well as General Electric Company personnel, came to i
the site to review the problem but nothing further was done about it.
Finding.
Inspection Report No. 50-293/76-04, which is a report of an inspection conducted by Region I at Pilgrim Unit 1 on February 13, 14, and 25, 1976, states that the licensee notified Region I by telephone on February 10, 1976, that twenty-one surface cracks were found in the interior cladding of one feedwater nozzle. The report contains additional information which was obtained during the inspection concerning the matter, including the licensee's planned corrective action. The report indicated the matter was considered unresolved I
pending a review of the corrective action during a future NRC inspection.
Inspection Report No. 50-293/76-05, which is a report on an inspection conducted by Region I at Pilgrim Unit I on March 15-18, 1976, states that an inspection was made of the grinding performed to remove the cracks and that observations were made of the resulting cavities and the performance of the nondestructive examinations.
It further states that an evaluation was being performed by the General Electric Company design engineers and that the status of the matter is still unresolved pending further review during a subsequent NRC inspection.
I180 102
15 Report No. 50-293/76-08, which is a report on an inspection conducted by Region I at Pilgrim Unit 1 on April 6-8 and 15, 1976, states that the General Electric Company had completed removal of the cracks, verified that the minimum wall thickness was maintained, and determined through analysis that the areas were acceptable to continue service without additional repair..The matter was resolved at that time.
A copy of pertinent portions of the above-mentioned reports are attached to this report as Exhibit D.
4.
Management Discussion On August 22, 1979, the investigation findings were discussed with the SRI representatives having an asterisk beside their name in the i
Persons Contacted section of this report. They were informed of the specific allegations made and that no items of noncompliance with NRC requirements had be identified.
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.;te: This document was typed verbatijn from a copy of the handwritten original.
I make the following statement; In late 74 or early 75 during the fiRC investigation of Sullivans charges of falsefee falsifisation against SWRI I was told by and of Dept 17 to create a document for Brunswick plant in fiorth Carolina.
I was told what to say on the document and to sign it.
I later saw this document in the files at Dept 17.
Pior to geen going to work for dept 17 and Southwest Research I did not have a high school education or GED on my e=ployment status.
I stated I did and was hired. About tuo weeks after being employed I found out it was a AEC and industrial requirement to have a high soschool education as a rdm minimu:n requirement for certification as a inspector.
I went to Mr.
' my supervisor at the. time and told him that because there were Federal Reg's goverening this that I was offering my resignation.
My R resignation was refused and I was told later to get a GED which I did on my own.
While 'at SWRI during a inspection of Turkey Point III I saw documentation of 5.W. examination of a weld and another companies examation of the same weld.
The other companies examatien had no recordablo indications.
S.W.c$a=ationhad any several recordable indications.
I was told by other emploe,es who had been to the plant the other company had falsified the entire baseline exam.
Dur4ng-the-inseye During a inservice inspection of Big Rock Point of the vessal ' head, the level II, who was suppose to be watching the secpe walked of during' the exam. My t=am mate told mo to watch it avon though I was not eartified tn da' so during the exam I saw somthing I had doubts about.
When the Level II came e :
i bact my partner and I prot.est=d his actions to pur team lesder. The icvel II had written of the weld as no recordable indications the matter was dropped.
During the testing of the Bailey Heter Babcock Wilcox cabinet for the flRC in 77 I was told to work with uncalibrated equipment and the strain gages out of place against the project plans.
My supervisor was aware of this.
During the test a fiRC member came down from Wash and did not check to see if the equipment was caled caled or the gages in placa saaording to the project
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l 1180 104 Exhibit A Page 1 of 2a
, _p I would also like th to add the document for Brunswick was a fa falsifacation and my supervisors including-
- approved, were aware of it and was one of the peep people who esid rae and were among the people who told me of the falsifaction at Big Reek Turkay Point III also and were team leaders at the time and tnchnican.
was a Level II At Big Rock was the team was rqy team leader.
was the level II and.
was try lovel I partner.
The document for Brunswick was a Hag Particle Mfg manufactures statement as to the sulfur & halogins its the material used for the examination and also stated it was mixed according to mfss recommendations.
e@loyed when these inspections took place.
4-de-net-he44 eve I was"Ct August 8., 1979 N I" All of the crossed out words were initialed as
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2 MEMORANDUM Felsruary 14, 1975 TO:
Filo FROM:
SULJECT:
Magnetic Particic Examination for Brunswick Steam Electric Plant Unita 1 and 2
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In performing the magnetic particic examinatian of the reactor pressure vessel closure head studs at Brunsvick Steam Electric Plant Units 1 and 2, J,
the mixture used was No. 20A WAT1;R IJ1S18ERSABLE MAGNAGLO. TMs mixture was made by mixing eight parts ci solutton #WA-2A WATER CON-DITIONING AGL*NT and 1 part Magn.igtu No. 14A DATCII No. 2LO64 Oil Dispe rs,hle Magriagio pa r gallung of deionie.cd wate r.
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1180 108
EXHIBIT D ATTACHED:
1.
IE Inspection Report No. 50-293/76-04 2.
IE Inspection Report No. 50-293/76-05 3.
IE Inspection Report No. 50-295/76-08 e
0 1180 109
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IE:I Form 12 l
(Jan 75) (Rev)
U. S. NUCLDR REGULATORY CO:PIISSION OFFICE OF INSPECTION AND ENFORCDENT l
REGION I l3i x
50-293/76-04 O
IE Inspection Report No:
Docket No:
50-293 Licensee:
Boston Edison Company License No: DPR-35 800 Boylston Street Priority:
Boston, Massachusetts 02199 Category:
C Safeguards Plymouth, Massachusetts (Pilgrim Unit 1)
E Location:
Type of Licensee:
BWR (664 de~T Type of inspecticn:
Incident, Announced Dates of Inspection:
February 13-14, 25, 1976 February 3-5, 1976 Dates of Previous Inspection:
Reporting Inspector:
h h
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9 /0 74 G. Walton, Reactor Inspector DATE '
Acco:panying Inspectoes:
None DATE DATE DATE J. Geiske and B. Nichols on Tebruary Other Acconpanying Personnel:
Sandia Laboratory 25, 1976 DATE
[,
Reviewed By:
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R. C. Haynes, M tion Chief, Engineering Support Section DATE 1180 110 4
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SU W.ARY OF FINDINGS Enforcement Action None License-Action on Previousiv Idencified Enforcement Iters Not Inspected Design Changes See ite=s 1 and 3 of Unusual occurrences below.
Unusual Occurrences 1.
Reactor Pressure Vessel Feedwater Nozzle Cladding Cracks The licensee notified the NRC Region I Office by telephone on February 10,1976 that 2.1 surface cracks, 1/16 to 1/4 inches l_ong, were found in the interior cladding of one feedwater noz=le.
The licensee's correctiva action plans are to inspect the cladding of all four feedwater no: les, remove any cladding cracks and install newly designed feedwater spargers which have been shown at a similar BUR facility to' minimize the potential for future cracks.
(Details, Paragraph 2) 2.
Ultrasonic Examination of Reactor Pressure vessel to Foz=le N2B Weld The licensee notified the NRC Region I Office by telephone on February 12, 1976 that an inservice inspection of the vessel-to-no Ele weld of noz=le N2B during the current refueling outage revealed that the maximum size of a flaw in the weld is different from that previously reported.
(Details, Paragraph 3) 3.
Reactor Recirculation System Bvoass Pining The licensee notified the NRC Region I Office by telephone on February 12, 1976 that inservice inspections of the four-inch diameter reactor recirculation system bypass piping welds indicated a possible crack in each of two welds. The licensee plans to per=anently remove the bypass piping. This design change has been implemented at similar BWR facilities where f'.
D 1180 111
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.2-it has been shown that this piping, which has been susceptible to weld cracks, was not required and could be removed from the design.
(Details, Paragraph 4)
The above three unusual occurrences are considered to be unresolved items pending a review of the licensee's corrective actions during subsequent NRC inspections.
Other Significant Findines None Management Interview t
The inspector discussed the inspection findings with Mr. J. Larson, NucAear Licensing Administrator, at the conclusion of the inspection.
i l
The inspecter stated that no items of noncompliance were identified during this inspection.
The inspector also stated that the three unusual occurrences, as identified in the Suc=ary of Findings section of this report, were considereno be unresolved pending review during subsequent NRC inspectiotis of the licensee's
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co=pleted corrective actions, i
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1180 112
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e's DETAILS l
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Persons Contacted Boston Edison Cocoan*
J. Smith, Station Manager F. Faculari, Maintenance Department Engineer E. Kearney, Operations Engineering Group Manager
.6 Smith, Maintenance Engineer' l
J. Larson, Nuclear Licensing Administrator - Operations Southwest Research Institute i
A. Whiting, Manager W. Flach, Manager of Field Services C. Ca=pbell, Senior Quality Assurance Engineer B. Baker, NDE Engineer TeledneMaterialsResea$c G. Speiring, Principal Engineer Sandia Laboratories J. Geiske, Engineer
)
Consultants to the Nuclear Regulatory B. Nichols, Engineer )
Cocsission 2.
Reactor Pressure Vessel Feedwater Nozzle Claddine Cracks During the current refueling outage, the licensee perfor=ed a visual inspection of the reactor feedwater sparger and a surface examination of the adjacent stainless steel cladding on the interior of one feedwater nozzle.
These inspections were perfor=ed as a preventive measure since feedwater sparger and cladcHTg damage had been observed at siailar BWR facilities.
The licensee's visual inspection of the sparger (which consists of four perforated, curved sections of six-inch diameter stain-less steel pipe located inside~tEs~ pressure vessel with a sparger section fitted to each of four feedwater nozzles) revealed no observable damage had been experienced by the sparger.
The surface examination of the feedwater nozzle cladding adjacent to one sparger section revealed 21 cracks (1/16 to 1/4 inch long) in the cladding at the upper, inner radius area of the feedwater nozzle.
Only part of the nozzle i
cladding was examined at this time and the remainder of this b
l'180 113
i 7
nozzle's cladding and the cladding of the remaining three feedwater nozzles is scheduled to be examined later during the outage.
The licensee has removed the existing feedwater sparger and will install a modified sparger.
The new sparger is to reduce the potential for damage to the nozzle cladding.
At the time of this inspection, the licensee was in the process of removing the fuel from the reactor vessel in order to 6
I facilitate the additional inspection of the feedwater nozzle cladding, the removal of any cladding cracks and the installa-tion of the modified feedwater sparger.
This item is considered to be unresolved pending completion of the licensee's corrective actions which will be reviewed during a subsequent NRC inspection.
3.
Ultrasonic Examination of Reactor Pressure Vessel to Nozzle N2B Weld During the current refueling outage, the licensee perfor=ed inservice inspections of two previously identified flaws in the vessel-to-nozzle welds of nozzles N2B and N4A.
These inspections were performed in accordance with the requirements stated in the NRC letter of July 19, 1974 to the licensee in that two independent inspection con. tractors, the Southwest Research Institute and the General. Electric Company, performed j
independent ultrasonic examinations of the flaw areas.
These examinations, using the[ha=e calibration block e= ployed during the 1974 refueling outage, indicated that the N4A nozzle weld flaw was essentially the same. However, the maximum through wall depth of the flaw in the N2B nozzle weld appeared to be greater than that 3Ypviously determined.
One significant difference in the testing method during the 1976 inspection was that the equipment used to mechanically inspect nozzle N2B allowed the bottom of the flaw to be defined with th.e mechanical equipment.
During the 1974 inspection, less accurate manual techniques were e= ployed i
to define the bottom of the flaw. As a result, the 1976
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mechanical inspection indicated that the flaw area measured in 1974 was not the area of maximum through wall dimension -
rather, the maximum through wall dimension was located about 50 away in a clockwise direction. The apparent flaw size at this azimuth (243 ) was about 1.3 inches through wall com-pared to the previously reported 0.55 inch through wall i
at the 238 azi=uth. Mechanical measurements of the flaw i
e 1180 114 MM
o at the 238 azimuth, using the original calibration block, i
revealed that the flaw size was the same (i.e., within the tolerance of the measuring technique).
Subsequently, the licensee obtained a new calibration block i
fabricated from a prolongation of actual vessel shell material.
This material is of a higher quality than the material from which the original calibration block was fabricated.
Using the new calibration block, which is acoustically similar e
to the material to be inspected as required by by the latest l
revisions to the ASME Code, the licensee fbund that the apparent flaw size in both the N4A and N2B ozzle velds is significantly smaller than that flaw size defined using the original calibration block This effect was ascribed to the fact that the old calibration block, fabricated from lower quality steel, has a greater signal attenuation tha'n
that of the reactor vessel material.
)
The licensee is continuing his evaluation of the aforementioned
~
apparent changes in flaw size to develop a better understanding of the test results. This evaluation will include the following:
~
a.
Deter 4"e if the flaw has propagated to a larger size flav.
~
b.
Deter-fae if_the_ change is due to test variables.
Some of the known var _1 ables are: change in calibration technique and test block material, amount of transducer overlap, curved transducer shoe versus flat shoe, different transducer although same style, dia=eter and slight angle change, ~ manual inspection versus mechanical inspection.
c.
Determine the indicated flaw size by using different test techniques.
For example, the present ASME Section XI BSPV Code states the terminal ends of a flaw shall be determined using 50 percent of the calibrated i
distance a=plitude curve as the flaw size; however, in addition to the 50 percent level, the licensee is using 20 percent of the distance amplitude curve to define the terminal ends of the flaw.
This latter data is obtained for infor=ation purposes as a c5h-servative, upper limit of the flaw size.
Upon completion of his evaluation, the licensee will submitareportoftheinspectionresyltsandevaluation of these results to the NRC for review by Licensing.
4 This item is unresolved pending acceptable findings from this l
review.
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Reactor Recirculation System Bvoass Piping During the current refueling outage, the licensee conducted ultrasonic examinations of the.yelds in the four-inch recircu-i lating_ system bypass lines to_v_e_r_1fy weld quality.
This exambation revealed linear indications in two welds.
The licensee perforned a radiographic examination of Ihe' welds for additional verification of the presence of linsar indications.
The licensee intends to co=plete a design change for removal and capping of the two four-inch recirculation system bypass lines.
This fix was employed recently at similar BWR facilities to eliminate the four-inch bypass lines which had experienced recurrent cracks.
This item is considered unresolved pending review during a subsequent inspection of the licensee's corrective actions.
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--f *k j Qan 75) (Rev)
, U. S. I.'UCLEAR REGULATORY CC:CJISSIO:i OFFICE OF I:;SPECTIO:? A::D E!?FORCD:E:Tr RECIO:t I ws.,
E Inspection Report !!o :
50-293/76-05 Docket Io:
50-293 @;
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Licensce:
Boston Edison Company License No: DPR-35 800 Boylston Stree
~~
Priority:
Boston, Massachusetts 02199 Category:
.C Safeguards Loca tion:
Plymouth, Massachusetts (Pilgrim Unit 1)
Type of Lic2nsco:
BWR (664 MWe)
Incident. Announced l
'fype of Inspection:
Dates of Inspection: March 15-18, 1976 Dates of Previous ' Inspection:
February 13-14, 1976 Reporting Inspector: '
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G. Walton, Reactor Inspector DATE Accompanying Inspectors:
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W. Sanders, Reactor Inspector DATE DATE DATE Other Accompanying Personnel:
None DATE Reviewed By:
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M/7.!76 R. C. diayned, Selt[on Chief, Engineering Support UAIE 1180 117
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t sum!ARY OF FINDINGS S*
Enforcement Action Items of Noncompliance None identified.
Licensee Action on Previously Identified Enforcement Items Not inspected.
Design Changes A.
Feedwater Sparger Replacement The feedwater sparger and thermal sleeves are being replaced with a design that utili=es a metal to metal interference fit between the feedwater nozzle ID and the thermal sleeve OD.
(Details, Paragraph 3)
B.
Reactor Vessel Recirculation System Byoass Pioing During this outage, the 4" OD bypass piping assembly around the 28" motor operated valve (MOV) will be removed and eliminated. The 4" lines are to be separated 8" from the recirculation line on the upstream side of the MOV and 5" from the recirculation line on the downstream side.
The stubs will be sealed by welded hemispherical caps.
(Details, Paragraph 4)
Unusual Occurrences A.
Feedvater Nozzle Cladding Cracks
Reference:
Inspection Report No. 76-04 An inspection was made of the grinding performed to remove the cracks found in the inside ladded radii of the feedwater nozzles.
Observations were mada of the resulting cavities and the performance of the nondestructive examinations.
(Details, Paragraph 2)
B.
Jet Pump Restrainer, Bolt Keeoer Weld Cracks A visual inspection performed by the licensee of integrally welded internal supports revealed cracks in the jet pump restrainer bolt keeper tack welds.
Corrective measures are oeing developed.
(Details, Paragraph 5) 1180 118
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Ultrasonic Examination of Reactor Pressure Vessel to Nozzle N23 Weld
Reference:
Inspection Report No. 76-04 The inspector attended a meeting at the home office of the licensee's inspection contractor, Southwest Research Institute, at San Antonio, Texas, to review the ultrasonic test data obtained from the 1974 and 1976 inservice examinations performed on no==le weld N2B.
(Details, Paragraph 6)
Other Sienificant Findings l
A.
Current Findings Acceptable Areas Procedures An inspection of the following items were perfor=ed:
Procedures for the removal of feedwater nozzle cracks.
(Details, 2)
Procedures for the replacement of feedwater spargers.
(Details 3)
Procedures for the modification of 4" bypass line.
(Details 4)
B.
Unresolved Items Items A,B, and C of the Unusual Occurrences are considered unresolved pending a review of the licensee's actions during subsequent NRC inspections.
C.
Status of Previousiv Reoorted Unresolved Items See paragraphs A and C of Unusual Occurrences.
Deviations None identified.
Management Interview i
At the conclusion of the inspection, a meeting was held at the site with the following people to discuss the inspectic 1180 119
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. Boston Edison Cocoany J. Nicholson, Chief, Maintenance Engineer J. McLaughlin, Compliance Engineer Purpose of the Insoection The inspector stated that the purpose of this inspection was to inspect the procedures and work in progress for the items listed.
A.
Feedwater Sparger Replacement.' (Details, Paragraph 3)
B.
Reactor Vessel Recirculation System Bypass Piping.
(Details, Paragraph 4)
C.
Feedwater Nozzle Cladding Cracks.
(Details, Paragraph 2)
D.
Jet Pump Restrainer, Bolt Keeper Weld Cracks.
(Details, Paragraph 5)
Within tha scope of this inspection no deficiencies were identified.
1180 120
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DETAILS t
1.
Persons Contacted Boston Edison Cocoany J. Smith, Pilgrim Station Manager S. Martin, Management Systen Coordinator J. Nicholson, Chief Maintenance' Engineer J. McLaughlin, Compliance Engineer D. Clark, Compliance Engineer F. Famulori, Maintenance Depart =ent Engineer R. Bohlman, Security Supervisor General Electric Company P. McGuirp, Resident Field Engineer J. Folk, Nondestructive Examiner Level III J. Tucker, Nondestructive Examiner Level II C. 3rookelman, Supervisor Installation and Service Engineering T. Maikoff, Supervisor Installation and Service Engineering Nuclear Plant Services H. Smith, Health Physics Supervisor Southwest Research Institute W. Flach, Manager, Field Services B. Baker, NDE Engineer W. McGaughey, Test Engineer Battelle, Pacific Northwest Laboratories G. Posakony, Manager, Nondestructive Testing, Consultant to Boston Edison Company Teledyne G. Spierling, Principal Engineer, Consultano to Boston Edison Company Oak Ridge National Laboratories K. K. Klindt, Test Engineer, Consultant to the Nuclear Regulatory Commission 1180 121
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- l Sandia Laboratories J. H. Gieske, Test Engineer, Consultant to the Nuclear Regulatory Cocmission 2.
Feedwater Nozzle Cladding Cracks The visual inspection of the reactor feedwater spargers and the inside cladded radii of the feedwater nozzles identified surface cracks in the cladding.
The results of this examination are described in IE Inspection Report No. 76-04.
The inspector observed the work in progress in the reactor vessel to remove the cracks and reviewed the procedures for performing the work.
The cracks are being removed by General Electric Company personnel using grinding procedures similar to those developed for the re-moval of cracks at the Monticello facility in late 1975. After each of the cracks are removed, the result cavity is examined with dye penetrant to verify the elimination of the defect.
The re-sultant cavity is also ground to a slope on all sides to reduce stresses. This work is performed in accordance with the Procedure FWSR 6.0 SI Revision 1, "...Feedwater Nozzle Blend Radii Grindout Etch Procedure...."
This procedure is used to etch the grindout cavity to establish the junction line between the carbon steel and the stainless steel.
This will then provide a defined line to accurately measure the exposed area and penetration into the carbon steel.
Procedure SPCS 6.0 5-1-1 is used to record and layout the location of the,grindout cavities to the nozzle axis.
In addition the inspector was inf.Ted that impressions of each cavity would be made using dental material.
During the removal of the cracks, it was found that the depth of at least one grindinc cavity had penetrated the base material to approximately 3/8".
The evaluation of the acceptability of the grindout cavities is being performed by General Electric Company design engineers.
1 This item is considered to be unresolved, pending a review of the licensee's action during a subsequent NRC inspection.
i180 i22
(
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. 3.
Feeduc' 2r Soarger Reolacement i
During this refueling outage, the feedwater spargers are being re-placed with a new design that has a 10 mil interference fit between the nozzle and the ther=al sleeve. This is obtained by machining a band on the outside diameter of the ther=al sleeve tg 10 mil over the
~ actual measured inside diameter of the nozzles.
The thermal sleeves are i=mersed in liquid nitrogen to " shrink" their diameter and then they are inserted in the nozzles. This sparger modification is similar to that at the Millstone 1 facility.
a The inspector inspected the work in progress, which consisted of the j
weld joint and consumable insert fit up of the first ther=al sleeve to spargar assembly.
In addition, the following procedures and qualifications were inspected for compliance to the contract require-I ments and to Section IX of the ASME Code for the qualification of I
welding:
t FWSR 1.0 - Revision 0 -
" Material and Processes." This proce-dure lists materials that can be used on BWR Systems.
FWSR 2.0 - Revision 0 -
" Reactor Cavity and Reactor Pressure l
Vessel Decontesination'.' Describes in detail the rteps to decontaminate the reactor cavi y using hydro laser equip-ment.
FWSR 4.0 - Revision 2 -
" Installation of In-Vessel Platform and Services."
FWSR 4.1 - Revision 1 -
" Reactor Feedwater Nozzle and Sparger Inspection."
Inspection r;Cgram for sparger assembly including liquid penetrant.
SPCS 4.1.1 -
" Maintenance Special Process Control Sheet." Visual inspection data sheets including defect map for assembly parts.
FWSR 5.0 - Revision 2 -
"Feedwater Sparger Removal." To remove
?
old spargers.
FWSR 7.0 - Revision 1 -
"As-Built Data Determination." Procedure for measuring to establish diameter of thermal sleeves for interference fit.
FWSR 7.1 - Pevision 0 -
,"Sparger Fit-Up " Procedure for the fit up and locations of the thermal sleeve and attachment brackets.
1180 i23
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3 FWSR 9.0 - Revisien 2 -
"Feedwater Sparger Replacement Proce-dure." Detail steps for installing the new feedwater sparger design.
FWSR 8.0 - Revision 1 -
" Welding of the Feedwater Spargers."
Details for field welding of the various i
pieces of the sparger assemblies.
4 Welding i
1SE-GWP-N - Revision 2 -
" General Welding Procedure for Nuclear Service"
~'
1SE-GWP-NSI - Revision 3 -
" Supplement to General Welding Procedere for Field Modification of Reactor Vessel Components."
1SE-DWP-3004 - Revision 3 -
" Weld Procedure Carbon Steel / Stainless Steel Gas Tungsten Arc Weld (GTAW) Consumable Insert 15E-PQ-306-2G Procedure Qualifications for the 1SE-PQ-307-6G above Procedure Nondestructive Examination 1SE-QAI-110 - Revision 3 -
" Visual Examination Test Procedure" This includes the requirement to be able to distinguish a 1/32" black line on a 18%
grey background.
ISE-QAI-330 - Revision 5 -
" Color Contrast Liquid Penetrant Procedure" 4.
Reactor Vessel Recirculation System Bynass Pining The ultrasonic examination performed by the licensee during the current refueling outage revealed the presence of linear indications in two of the welds.
Radiographic examinations have confirmed the indications.
A decision was made to modify the recirculation system by eliminating the 4" OD line which was originally designed and used as a bypass line around the 28" Motor Operated Valve (MOV) in the recirculation line.
This design modification has been made at other BWR facilities.
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The modification will separate the bypass line 8" from the recircu-lation line on the upstream side of the MOV and 5" from the recircu-lation line or the downstream side of the MOV. The remaining stubs will be sealed by welded hemispherical caps. Both the A and the B lines will be modified in this manner and is expected to be started by the end of this month.
The inspector reviewed the job procedures; no deficiencies were identified. The proceduren are,:
76-47 " Removal and Capping of Reactor Recirculation System 4" a.
Bypass Piping" b.
C-59-Revision 2
" Welding Procedure" t
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" Liquid Penetrant Test Procedure - Factory Mutual Research Corporation c.
Radiographic Test Procedures (RT)" - Factory Mutual Research Corporatio This procedure requires RT of the root pass and finish weld.
The leak test is performed with the Vessel Hydro Test.
5.
Jet Puen Restrainer, Bolt Keener Weld cracks A visual inspection performed by the licensee, of selected reactor vessel interior surfaces with integrally welded internal supports, in accordance with the requirements of PNPS Technical Specification 4.6.C, revealed cracked tack welds on the jet pump restrainer bolt keepers.
The inspector ascertained that the corrective actions are being determined by General Electric Company and were initiated by issuing a FDI to detension two (2) of the top hold down bolts for measurements.
The information obtained will be forwarded to General Electric Company Engineering for evaluation and deternina-tion sf additional actions.
The item is considered to be unresolved, pending a review of the licensee's action during a subsequent NRC inspection.
6.
Ultrasonic Examination of Reactor Pressure Vessel to Nozzle N2B Weld On March 17-18 the inspector, accompanied by two consultants, visited the home office of the licensee's inspection contractors Southwest Research Institute (SWRI), San Antonio, Texas, for the purpose of obtaining additional data on the nozzle weld N2B.
I Additional examination analysis was performed by SWRI on the two ultrasonic calibration blocks to determine the active beam spread using the ASME, B&PV Section XI 50 percent distance amplitude points 1180 125
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. t to determine terminal ends. The beam spread measurements indicate that actual defect measurements using the calibration obtained from the original calibration block (PIL-5) would magnify the defect size and error could occur.
In addition to beam spread, the ultrasonic sensitivity differences between the two calibration blocks, (PIL-5, PIL-5A) were evaluated.
The difference between the two blocks is a factor of 5 with the PIL-5A giving the smaller response.
When using the original calibra-tion block PIL-5, this would also cause the defect in the nozzle weld N2B to be magnified and oversized.
The inspector reviewed the mill certification for the calibration block PIL-5A to determine the origin of the material.
The records indicate the PIL-5A material, heat nu=ber C2945-2 is a prolongation j
from the Pilgrin 1 reactor vessel.
The licensee has performed an evaluation using the correct calibration block PlL-5A and determined the defect in nozzle weld N2B is substan-tially smaller in through-wall dimension than the 1.3 inches that the original test indicated.
The inspector and consultants obtained reproduced strip chart readings of several examinations performed during the 1974 refueling outage and the 1976 refueling outage.
This data will be evaluated by the two consultants to (a) determine if the flaw has propagated during reactor operation, and (b) determine the defect size using beam spread data, different scans on both calibration blocks and different amplitude responses.
This item is considered to be unresolved, pending review of the consultant's evaluations.
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32: T Form 12 C.:n 7.5) Gev)
' U. S.1:UCLEAR REGU!ATORY CC:CIISSION OFFICE OF I:;S?ECTIO: A::D E::FORCD:E:.T REGIO:: I IE Inspection Ecport No:
50-201/76-08 Docket Nc:
_50-293 Licensce:
Boston Edison Codoany License No: DPR-35 800 Boylston Street Priority:
Boston, Massachusetts 02199 Category:
C Safeguards
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Group:
.y.3 Loca tion:
P1vmouth. Massachusetts (Pilerin Uair IT g
y -w Type of Licensee:
BWR (664 MWe) m Type of Inspection:
Incident, Announced Dates of Inspection:
April 6-8, and 15, 1976 March 15-18, 1976 Dates of Previous Inspection:
Reperting Inspector:
I h,h f/d 1/ 74_
G.
A.
Walton, Reactor Inspector DATE Accompanying Inspectors: c/// [S o d,. ]
d'~/2e/7t, F.
e W.
nders, Reactor Inspector
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DATE den)-
fxhs W.
Devlin, Investigator DATE DATE Other Acconpanying Personnel:
None DATE Reviewed By:
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C.
- Haynes, Chid", Engineering Support Section 1180 127 3
i SUmfARY OF FINDINGS Enforcecent Action None.
Licensee Action on Previous 1v Identified Enforcement Items Not inspected.
Desien Changes A.
Feedwater Soareer Replacement
Reference:
IE Inspection Report 50-293/76-05 The feedwater sparger and thermal sleeves have been replaced with a design that utilizes a metal-to-metal interference fit between the feedwater nozzle and thermal sleeve.
This item is considered resolved.
(Details, Paragraph 2)
B.
Reactor Recirculation System Bypass Piping
Reference:
IE Inspection Report 50-293/76+05 The piping was revised by eliminating the 4 inch bypass piping around the 28 inch loop discharge valves.
This item is considered resolved.
(Details, Paragraph 3)
Unusual Occurrences A.
Feedwater Nozzle Cladding Cracks
Reference:
IE Inspection Report 50-293/76-04 and; IE Inspection Report 50-293/76-05 The licensee has removed by grinding the cracks found in the cladding on the inner radii of the reactor pressure vessel feedwater nozzles.
This item is considered resolved.
(Details, Paragraph 4)
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B.
Jet Puca Restrainer, Bolt Keeuer Tack We?.d Cracks
Reference:
IE Inspection Raport 50-293/76-05 A visual inspection performed by the licensee of integrally welded intcrnal supports revealed cracks in the jet pump restrainer bolt keeper tack welds.
Corrective measures were taken by the licensee.
This item is considered resolved.
(Details, Paragraph 5)
C.
Ultrasonic Examination of Reactor Vessel-to-Nozzle N2B Weld
Reference:
IE !aspection Report 50-293/76-04 and; IE Inspection Report 50-293/76-05 The licensee has perfer=ed additional examinations utilizing the Automatic Data Acquisition Systam and determined the defect in nozzle weld N23,when dimensioned by a circumscribed rectangle,10 approximately 0.9 inch in through wall dimensions.
This item continues to be unresolved.
(Details, Paragraph 6)
D.
Interview of Nondestructive Examination Personnel The investigative inspector and the inspector interviewed several Southwest Research Inntitute (SWRI) NDE technicians regarding state =ents made by one technician concerning the quality of some ultrasonic examinations perfor=ed during the current outage.
This item is considered to be resolved.
(Details, Paragraph 9)
Other Sienificant Findings A.
Current Findings Unresolved Items 1.
Inservice Inspection The inspector audited the licensee's inservice inspection program scheduled for completion during the present outage.
The inspector also discussed the licensee findings regarding the retest of certain welds which the licensee had committed to retest during this outage.
(Details, Paragraph 7)
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Acoustic Emission The licensee plans to monitor the behavior of the defects in RPV nozzle welds N23 and N4A using acoustic emissions.
(Details, Paragraph 8)
B.
Status of Previously Reported Unresolved Items
~~See_paragraohsAandBof"D7signChagesanFp5r5graphA and 3 of Unusual Occurrences.
C.
Deviations None identified.
Managenent Interview At the conclusion of the inspection conducted April 6 through 8,a meeting was held at the site with the following personnel to discuss the inspection.
Boston Edisen Ccepany J. Smith, Pilgri= Station Manager J. Nicholson, Chief Maintenance Engineer E. Kearney, Operations Engineering Manager F. Fa=ulari, Maintenance Department Engineer M. McLaughlin, Compliance Engineer W. Smith, Maintenance Engineer At the conclusion of the inspection conducted April 15,a meeting was held at the site with the following people to discuss the inspection.
Boston Edison Company J. Smith, Pilgrim Station Manager W. Smith, Maintenance Engineer E. Kearney, Operations Engineering Manager F. Faculari, Maintenance Department Engineer
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Southwest Resear~ch Institute M. Wise, Quality Assurance Manager Purpose of the Inspection The inspector stated that the purpose of the inspection wr., to inspect the work activities, procedures and progress for the items listed.
A.
Feedwater Sparger Replacement.
B.
Reactor Recirculation System Bypass Piping.
C.
Feedwater Noz::le Cladding Cracks.
D.
Jet Pump Restrainer, Bolt Kc.eper Tack Weld Cracks.
E.
Inservice Inspections.
In addition the inspector interviewed certain Southwest Research Institute nondestructive examination personnel.
Within the scope of this inspection no deficiencies were identified.
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DETAILS 1.
Persons Contacted Boston Edison Cocoany J. Smith, Pilgrin Station Manager S. Martin, Management System Coordinator J. Nicholson, Chief Maintenance Engineer
' i. McLaughlin, Compliance Engineer
~~~I. Faculari, Maintenance Department Engineer E. Kearney, Operations Engineering Manager W. Smith, Maintenance Engineer General Electric Cocoany V. Bain, Quality Control Engineer H. Erickson, Senior Engineer Southwes: Research Institute M. Wise, Quality Assurance Manager P. Ca=pbell, Senior Engineering Technologist J. Godwin, Crew Leader J. Johanson, NDE Technician W..'odell, NDE Technician Crouse Company B. Jones, Engineer 2.
Feedwater Soarger Replacement
Reference:
IE Inspection Report 50-293/76-05 The feedwater sparger design change which reduced the clearance between the thermal sleeve and the nozrle ID to a metal-to-metal interference fit has been completed.
The procedures, records, and work in progress were reviewed and reported in the referenced inspection report.
No deficiencies were identified.
This item is considered resolved.
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3.
Reactor Recirculation System Bvpass Piping
Reference:
IE Inspection Report 50-293/76-05 The action taken to modify the recirculation system by eliminating the 4" bypass piping around the 28 inch loop discharge valve in each of the two loops has been completed.
The program and procedures were reviewed and described in the referenced report.
}
After the removal of the bypass piping, the area which contained i
defects (detected by ultrasonic examination and confirmed by radio-graphy on previous tests) was cut to permit a dye penetrant test.
The defects were as follows:
a.
Weld No. 9, "B" Loop, Crane 4" Schedule 80 Elbow, 304SS, Heat No. C-2713 i
l First defect: a crack open from inside wall approximately 3/4 to 1" long, about 50% through wall, running circumferential and spaced approxi=ately 5/8" from fusion line of weld in the elbcw material.
Second defect: a crack open from inside wall approximately 1/2" long, about 30% through wall, running circumferential and located in middle of weld at the root.
b.
Weld No. 8 "A" Loop, Crane 4" Schedule 80 Elbow, 304SS, i
Heat No. C-2713 First defect: not visible, was shown by ultrasonic examina-i tion as a linear indication 3/8" - 1/2" long in weld.
i Second defect: not visible, was shown by ultrasonic examina-tion as a small pit in the center of the weld.
The licensee has made arrangements with the General Electric Company to perform metallogre.phic tests as a part of the continuing program for investigation of BWR pipe cracks.
The inspector had no further questions about this item at this time.
This item is considered resolved.
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4.
Feedwater No==le Cladding Cracks The licensee's contractor (the General Electric Company) completed the remcval of cracks,in the feedwater nozzle inner radius areas, which had been detected during liquid penetrant inspections.
(The reporting of these cracks and their re= oval by gri
'4.ng is described in IE Inspection Report Nos. 50-293/76-04 and 05.)
After removal of a crack, the resulting cavity was ground round-bottomed with a cdnimum radius of two times the depth of the cavity.
Subsequent to the grinding, the licensee's contractor made a chart of those indications which extended 3/16 inch or greater into base material. Twenty-five indications were charted and the maximum cavity extended into the base material 1/2 inch.
The contractor performed an analysis and determined the areas are acceptable for continued service without additional repair. He also verified that the minimum wall thickness was =aintained.
A final report will be submitted by the licensee to the NRC in accordance with his reporting requirements. The inspector had no further questions on this item.
This matter is considered to be resolved.
5.
Jet Fu=m Restrainer Bolt Keeper Tack Welds
Reference:
IE Inspection Report 50-293/76-05 The corrective measures developed by the designer (General Electric Company) to restore the keeper tack welds for the jet purp restrainer bolts and actions to preclude a recurrence of the failed tack welds have been co=pleted.
The licensee stated that the corrective actions included a change in two design requirements listed in the FSAR Figure 3.3.5, Notes 2 and 4; namely, an increase in the size of the tack welds to raise the break-away torque to 300 Ft-Lb on the two top hold down bolts and 100 - 200 Ft-Lb on the restrainer bolts.
The inspector was infor=ed that the corrective actions described were similar to the work perfor=ed at two other plants.
The inspector had no further questions and this item is considered to be resolved.
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6.
Ultrasonic Examination of Reactor Vessel-to-Nozzle N2B Weld Details of this item are included in IE Inspection Report 50-293/76-04 and 05.
The licensee performed additional inspections using ultrasonics on nozzle weld N23.
His evaluation of the data was that the =arfrum through wall dimension of the flaw was approxf=ately 0.5 inches.
However, the licensee reported that the orientation of the flaw was such that to dimension the flaw in the manner described by the ASME B&PV Code Section XI (circumscribed rectangle), the maximum dimension of the flaw was 0.9 inches through wall.
4 The licensee conducted an analysis to determine the maxi =um allow-able flaw size as required by the ASME B&PV Code,Section XI.
This analysis showed that the maximum allowable through wall dimension in the defect location was in excess of 2.5 inches.
The licensee submitted a report to the NRC on this matter which included the results of the analysis.
This item remains unresolved pending completion of the review of the licensee's report by the NRC staff.
7.
Inservice Inscection The inspector audited the licensee's inservice inspection' activities being perfor=ed during the current refueling outage.
The following areas were inspected:
Quality assurance inspections conducted by SwRI quality assur-a.
ance personnel.
b.
Personnel qualifications of SwRI test personnel.
c.
Review of ultrasonic calibration records.
d.
Review of customer notification forms (CNF).
Review of the scope of the inspection for compliance with the e.
licensee's Technical Specifications and the applicable ASME B&PV Code,Section XI.
1180 135
}
. During the current refueling outage, SwRI performed a retest of 38 welds in accordance with the licensee's coesitment to the NRC.
This co=mitment was a part of the licensee's planned action as a result of an allegation by a for=er SwRI employee who had participated in the 1974 inservice inspection.
(
Reference:
IE Investigation Report 50-293/74-01.)
SwRI cocpleted the retest and confirced that all but five welds gave ultrasonic responses similar, within 6 decibels (dbs), to the 1974 inspection and the preservice results. Of these five velds, three had signal responses as a result of the material geometry and were associated with the fabrication of the material.
The licensee established that these welds were acceptable without additional inspection.
The results of the tests of the remaining two welds were signifi-cantly different from the preservice and 1974 inservice inspection data. These welds (23-0-9 and 23-0-10) are in the HPCI steam supply piping located inside the drywell. During preservice inspec-tion the signal responses were xelatively low amplitude, 50 to 150 percent distance a:plitude curve (DAC), from root and crown reflec-tors.
In 1971, there were no recordable signals; however, during the recent reexa=ination, there were several indications at high amplitudes, up to 355 percent DAC.
These signals were of a magnitude that further evaluation was warranted.
The licensee contracted a third party inspection group (Factory Mutual) to radiograph the two welds to assist in the evaluation.
The radiographs confirmed the presence of an inside surface geometry condition on both welds.
In addition, weld 23-0-10 contained a scall crater crack. A review of the fabrication radiograph also confirmed the presence of the crack. The licensee reported this condition to the NRC and stated a repair would be perfor=ed.
The radiograph of weld 23-0-9 confirmed an inside root condition which was acceptable.
The inspector reviewed the aforementioned data with management personnel from BECo and SwRI.
The following additional infor=ation was provided.
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SwRI and BECo stated that the crater crack on weld 23-0-10 was of such a small magnitude that the ISI test in 1974 or 1976 would not reject this weld.
The inspector reviewed the original fabrication radiographs on veld 23-0-10 to assess whether additional review of other fabrication radiographs was warranted.
This review indicated that all code requirements for the radiograph were met, i.e.,
density, sensitivity, penetrameter, etc.
The crater crack could be confirmed only by the use of optical aids and the crack was quite small. The inspector stated no additional review of fabrication radiographs was necessary.
b.
The licensee stated that since positive correlation was not attained on all of the 38 welds reexamined, that BEco would retest all welds inspected by the three member team involved in the 1974 allegation.
This retest would involve of 34 2dditional welds and the licensee committed to retest as many as possible during the current outage. Any limitations on not completing the retest would be manpower availability and radiation exposure levels. The licensee stated that any of the 34 welds not retested during this outage would be completed during the next scheduled refueling outage.
The inspector stated the retest of the 34 welds was an acceptable program.
The inspector also stated this item would remain unresolved pending completion of the repair and completion of the retest program.
8.
Acoustic Emission
~
During the system hydrostatic test prior to resu=ption of power operation, the licensee will perform acoustic emission monitoring of previously identified flaws in the vessel-to-nozzle welds of nozzles N23 and N4A.
This requirement is stated in the NRC letter of July 19, 1974 to the licensee.
e I180 137
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Exxon Nuclear Company, Inc., through its NDT-Services Section, was selected by the licensee to conduct this monitoring and to evaluate the data.
This item is unresolved pending completion of the examination and a review of the test results during a subsequent NRC inspection.
9.
Interview of Nondestructive Examination Personnel On April 13, 1976, the NRC Region I Office was notified via telephone by SwRI management personnel that a SwRI technician currently at the Pilgrim 1 site had questioned the adequacy of ISI tests perfor=ed by a co-worker. The management personnel stated that this matter was confined to the 34 welds retested during the current outage (see Details, Paragraph 7). Boston Edison manage =ent also notified the NRC Region I Office of this matter via telephone on April 13, 1976.
The infor=ation received was that the employee told his supervisor that a co-worker had not performed certain test steps as required by the procedure.
Specifically, closecut calibration was not perforced for one weld and several welds were scanned too quickly.
SwRI management stated they had obtained written state =ents from both persons involved and these were available at the site. They also stated all personnel involved would be at the Pilgrim site until April 16.
They further stated that to remove any doubt, all exaninations (18 welds) performed by this team were being reexamined by other qualified technicians.
On April 15, 1976, an NRC Region I, Section Leader and a Reactor Inspector arrived at the Pilgrim site and met with BECo personnel and the SwRI Quality Assurance Manager.
The inspectors interviewed the SwRI Senior Engineering Technologist, the Quality Assurance Representative on site, the crew leader of the two workers involved in the dispute, the Level I Technician who had stated his concerns about the quality of work end the Level II Technician who had participated in the examinations.
In addition, discussions and records review was performed with the SwRI Quality Assurance Manager and the Maintenance Engineer of BECo who is responsible for the inservice inspection activities.
As a result of these interviews, and review of records, the inspector found no substantive deficiencies in the work as originally performed.
The retest of the 18 welds by other qualified technicians also con-fir =ed the acceptability of those welds.
This item is closed.
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