IR 05000280/1985015
| ML18142A492 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/22/1985 |
| From: | Blake J, Girard E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18142A489 | List: |
| References | |
| 50-280-85-15, 50-281-85-15, NUDOCS 8506270721 | |
| Download: ML18142A492 (12) | |
Text
Report Nos.:
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, ATLANTA, GEORGIA 30303 50-280/85-15 and 50-281/85-15 Licensee:
Virginia Electric and Power Company Richmond, VA 23261 Docket Nos.:
50-280 and 50-281 License Nos.:
DPR-32 and DPR-37 Facility Name:
Surry 1 and 2 Inspection Conducted:
22 - 26, 1985 Approved
/Datd Signed
~~Jr ate Signed SUMMARY Scope:
This routine, announced inspection entailed 40 inspector-hours on site in the areas of cracking in steam generator girth welds, inservice inspection, review of information related to proposed Technical Specification changes, and inspector followup item Results:
No violations or deviations were identified *
,~--8506270721 850528 PDR ADOCK 05000280
PDR-
REPORT DETAILS Persons Contacted Licensee Employees
- R. F. Saunders, Station Manager
- E. S. Grecheck, Superintendent of Technical Services
- J.M. McAvoy, Systems Engineer
- E.W. Throckmorton, Director of NOE (Nondestructive Examination) Services
- R. F. Driscoll, Manager of QA (Quality Assurance)
H. Miller, Assistant Station Manager, Nuclear Safety and Licensing R. Blount, Supervisor of Performance Engineering A. McNeil, Engineer E. Holloway, NOE Coordinator D. Spooner, Level III NOE Examiner Other licensee employees contacted included construction craftsmen, engineers, technicians, operators, mechanics, security force members, and office personne Other Organizations M. Diehl, Inspector, The Hartford Steam Boiler and Insurance Company D. Kurek, Level III Examiner - Inservice Inspection, Westinghouse Electric Corporation S. Reynolds, Reactor Inspector, NRC Region I D. Smith, Metallurgical Engineer, NRC Office of Nuclear Reactor Regulation C. Czajkowski, Investigator, Brookhaven National Laboratory NRC Resident Inspectors D. J. Burke, Senior Resident Inspector
- M. J. Davis, Resident Inspector
- Attended exit interview Exit Interview The inspection scope and findings were summarized on April 26, 1985, with those persons indicated in paragraph 1 abov The inspector described the areas inspected and discussed in detail the inspection finding The inspector expressed his concern that the licensee has no plans to inspect steam generator (SG) girth weld #11 for cracking even though it lies in close proximity to another SG girth weld that has experienced cracking that was apparently environmentally induce The inspector noted that outside diameter (OD) ultrasonic examinations for such cracks would require a very limited expenditure of time and radiation exposure, as the weld is currently readily accessible for such an examinatio i,"*
-
The following new items were identified during this inspection:
Inspector Followup Item,280, 281/85-15-01:
Future inservice inspection and repairs for steam generator girth weld cracking, paragraph Unresolved Item 280, 281/85-15-02:
Adequacy of transducer for ultrasonic examination of cast stainless steel piping, paragraph 6.a.(2).
Inspector Followup Item 280, 281/85-15-03:
Verification of augmented inservice inspections, paragraph The licensee did not identify as proprietary any of the material provided to or reviewed by the inspector during this inspectio.
Licensee Action on Previous Enforcement Matters This subject was not addressed in the inspectio.
Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations or devia-tions. A new unresolved item identified during this inspection is discussed in paragraph 6.a.(2). Cracking in Steam Generator Girth Welds (927018)
Note:
This subject could be addressed and covered as inservice inspec-tion (paragraph 6 below) or, even more appropriately, as Inspector followup item 280, 281/83-19-02 (paragraph 8).
However, because of its importance and the extensive inspection efforts involved, it is being treated as a separate inspection are References:
Letter from J. C. Miller (Westinghouse Electric Corporation)
to W. L. Stewart (Virginia Electric and Power Company) dated 6/10/82, subject - request for 100% ultrasonic examination of a steam generator upper shell/cone weld to determine whether
. cracking was present as previously (4/82) identified at the Indian Point 3 nuclear plan IE (Inspection and Enforcement) Information Notice No. 82-37:
Cracking in the Upper Shell to Transition Cone Girth Weld of a Steam Generator at an Operating Pressurized Water Reactor NRC Inspection Report 280, 281/83-19, dated 9/1/83 NRC Inspection Report 280, 281/83-22, dated 9/22/83 Report NUREG/CR-3281, BNL-NUREG-51670, Investigation of Cracking on the Steam Generators at Indian Point Unit No. 3, prepared by C. J. Czajkowski, published June 1983
3 Introduction The licensee has identified cracks in the upper shell to transition cone welds of their Unit 2 steam generators (SGs).
The NRC inspector examined the licensee's actions in addressing this matter through a review of the data recorded and through discuss ions with cognizant personne The inspector's examination was conducted to evaluate the adequacy of the licensee's actions with regard to regulatory require-ments and to obtain technical information that could be useful in evaluating the generic implications of the crackin Background information on this matter an~-a description of the inspector's exam-ination and findings are provided belo Background During the 1983 Unit 2 refueling outage, the licensee ultrasonically examined the upper shell to transition cone girth weld (weld 1-6) on SG The steam generator weld examination, in part, addressed concerns for potential cracks in such welds identified in IE Information Notice 82-37 (dated 9/16/82) and in a letter from Westinghouse to the licensee (dated 6/10/82).
These concerns resulted from cracks found in simi-larly located steam generator welds at the Indian Point 3 nuclear plan The licensee's examination of the Surry 2 steam generator weld found significant reflectors with amplitudes of up to 200% of DAC (the distance - amplitude curve).
The reflectors were at the steam gene-rator ID and lay parallel to the wel The reflectors were detectable from both sides of the weld, but the greatest amplitudes were observed in scans directed from below the wel The reflectors were not con-tinuous by ASME standards, but it appeared that the reflections were produced by a si"ngle continuous reflector if the examination was performed with sufficient gai The licensee's ultrasonic examination (UT) of the weld was observed by the NRC inspector and was documented in NRC Inspection Report 280, 281/83-1 The licensee had not completed their evaluation by the end of the 83-19 inspectio At the conclusion of that inspection, the inspector opened an item for followup inspection of the licensee's evaluatio The item was opened because the inspector was concerned that the adequacy of the licensee's evaluation of the reflectors be further verifie The followup item, identified as 280, 281/83-19-02, also addressed the licensee's examination and evaluation of reflectors previously revealed and accepted by the licensee in a similarly located Unit 1 SG wel In NRC inspection 280, 281/83-22, conducted two weeks after the 83-19 inspection, NRC Region II inspectors addressed the followup item and conducted an independent ultrasonic sample (approximately 5% of the length) examination of Unit 2, SG A, weld 1-6 *. At the time of the
inspection, the licensee had concluded their evaluation and had attri-buted the reflectors in the weld to acceptable weld fabrication defects and inside surface geometr The licensee's disposition of the reflec-tors had been based on location plots, radiographic examination data, visual fiber optic examination of the weld ID and holographic UT of a portion of the wel Based on the independent UT conducted during NRC Inspection 83-22, the NRC inspectors concluded that at some points the reflectors appeared more cracklike than geometric and it was not conclusive whether the signals were due to ID geometry or short shallow cracks. If they were cracks it was considered that they would not be a problem for the short ter The similarly located welds on Unit 2 steam generators B and C were not examined by eith'er the 1 icensee or the NRC inspectors. Based on the NRC inspectors findings, the licensee agreed to remove a section of the SG A Downcomer Resistance Plate and inspect a representative sample of the indications by magnetic particle (MT) inspection on the inside surface at the next refueling outag The next refueling outage was underway at the time of the NRC inspec-tion covered by this repor Prior to the start of the NRC inspection the licensee had conducted the agreed to MT inspection and had detected cracking at the SG A upper shell to transition cone wel Examination and Findings (1) Licensee Investigation of Cracking Discussions with involved licensee personnel disclosed that they and their contractor, Westinghouse, had performed the following investigative work:
(a)
MT and visual inspections of portions of weld 1-6 after polishing to remove corrosion products (photographic records were made)
(b)
UT examination of weld 1-6 for crack depth determination and for changes since the 1983 examination (c)
Local ID grinding and mechanical measurements on weld 1-6 cracking, followed by MT inspections to determine crack depth for correlation with UT data (d)
UT of the similarly located welds 2-6 and 3-6 on SGs Band C, respectively, for indications of cracking (e)
Review of radiography records for weld 1-6 Note:
Records for welds 2-6 and 3-6 could not be located (f) Review of design and operational data for possible relation to the cracking
(2) Technical Findings A summary of technical findings based on the inspector's review of related data and discussions with knowledgeable personnel is as follows:
(a) The Surry Unit 2 SG A upper shell to transition cone weld (weld 1-6) exhibited ID crackin (Information source -
photographs and discussions with licensee personnel)
(b)
The cracking is discontinuous surface cracking, apparently running around almost the entire circumference of the vessel, and is so closely spaced as to be considered continuous in accordance with the applicable code - ASME Section XI (80W80).
(Information source - photographs and discussions with licensee personnel)
(c)
The cracking lies above the transition cone to upper shell weld approximately at the line of the fusion but in the heat affected zone of the base meta (Information source -
photographs and discussions with licensee personnel)
(d)
The cracking appears shallow by comparison to the vessel wa 11 thickness of over 3 1/2 inche It appears to average about 1/8 inch deep with maximum depths estimated up to 5/16 inc Better estimates of crack depths will be available when the licensee performs anticipated grinding repairs to remove the crackin (Information source - review of UT data and discussions with licensee personnel)
(e)
Pitting is visually apparent in the weld and adjacent base meta The concentration of pitting is significantly greater above the weld - the pitting density a foot above the weld appears as great or greater than the density at or just above the wel (Information source - discussions with licensee personnel)
(f) At least some of the cracks run from pit to pi (Informa-tion source - discussions with licensee personnel)
{g) The cracks are not visible to the unaided eye but are visible (after removal of corrosion products) with the aid of a lOX magnifying glas ( Information source - discussions with licensee personnel)
(h)
Visually, the corrosion product at the ID weld area appears to be black magnetit (Information source - discussions with licensee personnel)
(3)
(i)
UT data, as interpreted by the NRC inspector, indicates that the average depth of the crack has increased since the 1983 refueling outag There does not appear to be any large increase in maximum depth, howeve (Information source - UT reports for 1983 and 1985 examinations of weld 1-6)
{j) The cracking is similar to that observed at similarly located welds in the Indian Point 3 steam generator (Information source - discussions with NRC and national laboratory person-nel who investigated the Indian Point 3 cracking)
(k)
The cracking is believed to have been induced by the local water chemistry, fabrication stresses or cyl i c thermal stresses or some combination thereo Post weld heat treat-ment temperature may also have been a factor - it is sug-gested that the temperature achieved may have been inadequate to properly temper the martensitic structure in the heat affected zone of the wel (Information source - discussions with licensee personnel and with NRC and national laboratory personnel who investigated the Indian Point 3-cracking)
{l) Unit 2 SGs Band C appear to have similar but less severe cracking at their transition cone to upper shell weld (Information source - UT reports for partial examination of SG Band C during the 1985 refueling outage)
Note:
Subsequent to completion of this NRC inspection, the licensee confirmed similar but apparently less severe crack-ing detected by MT inspection of the weld on SG ~ - This was described to NRC Region II in a telephone conversation on 4/29/8 (m)
Weld 1-6 is the original closure weld in SG As such it was in service for over ten years and was exposed to signifi-cant levels of chloride and possible copper intrusion from condensor tube leaks. (Information source - discussions with licensee personnel)
(n)
The presence of copper and chloride ions was implicated in the cracking that occurred in the Indian Point 3 SGs and has been demonstrated to accelerate cracking in laboratory tests on SG materials. (Information source - investigator from Brookhaven National Laboratory)
Licensee's Corrective Action The NRC inspector questioned licensee personnel as to the plans for the SG A cracking already identified and for further inspec-tion for SG weld cracking during the current outag The inspec-tor was informed that the licensee planned to remove the cracking identified in SG A by grinding and that they planned MT inspection
to determine the presence of cracking in the similarly located weld in SG They indicated that further inspection and/or repair to the transition cone to upper shell welds in SG Band C would be determined after they had obtained information from the MT inspection of the SG B wel *
The inspector questioned licensee personnel as to whether they were planning to inspect the SG A weld 1-11 which lies about six inches below and parallel to the cracked weld 1-They stated they did not intend to inspect weld 1-1 Reasons provided by the licensee were as follows:
Weld 1-6 was the original SG A closure weld that had experi-enced over ten years of service including several in which water chemistry had been unsatisfactory, largely due to SG tube leak Weld 1-11 was a new weld, the closure weld performed at the replacement of SG A, and it had only expe-rienced good water chemistr Weld 1-11 experienced less severe thermal stresses due to desig Weld 1-11, unlike weld 1-6, does not lie at a gross structu-ral discontinuit The design stresses are therefore lower and it is not required to be examined by the applicable cod UT of weld 1-11 might reveal non-relevent indications that would require evaluation and delay completion of the refuel-ing outag If cracking did occur the ductility of the vessel steel would result in a detectable leak well in advance of a weld fail-ur The inspector agreed with the licensee's reasonin He stated however, that while he also believed that the probability of significant cracking at weld 1-11 was low, he believed it should be ultrasonically examined in at leiist a sample area for the following reasons:
Location or environment appears to be a significant factor in the cracking, as similarly located welds exhibited cracking in Indian Point 3 steam generators Surry Unit 2 Weld 1-11 is located relatively close to the weld location that exhibit cracking - about six inches (less than two wall thicknesses)
from weld 1-6.
Currently, access to and examination of a portion of weld 1-11 would require only a few man hours and relatively low radiation exposure (a few tens of millirems)
The licensee has no plans to ever examine weld 1-11 during the life of the plan ~-
It was reported to the inspector that one of the examiners who performed UT on weld 1-6 had scanned down on weld 1-11 and observed significant indication Note:
Licensee personnel subsequently informed the inspector that they had questioned the examiners about this and were told that the examiner had seen indications but that they were not signif-icant in that they had been detected only at a very high gain settin The cause and time of initiation of the cracking and the rate of growth are not definitely established for Surry Unit A significant portion of weld 1-11 was examined prior to service and records indicated no indications above 50% of DA In addition to his concern for examination of weld 1-11, the inspector indicated continued general concern and interest in regard to the licensee 1s further examination and repairs related to the SG crackin The inspector informed the licensee that, because of this concern and interest, a followup item would be opened for further examination of the licensee 1s actions in subsequent NRC inspectio The item was identified as Inspector followup item 280, 281/85-15-01, Future Inservice ~nspection and Repairs for Steam Generator Girth Weld Crackin Matters which will be addressed in the followup will include:
The need for additional licensee examinations and data reviews to locate other cracking as in Unit 2 weld 1-11, Unit 1 SG transition cone to upper shell welds, feedwater nozzles, et The adequacy of repairs of the SG cracking The adequacy of examinations performed to verify removal of cracking The frequency of subsequent inspections on the areas that have experienced cracking Within the areas examined, no violations or deviations were identifie.
Inservice Inspection (ISI) - Unit 2 The inspector reviewed ISI procedures and observed ISI work as described below to verify compliance with licensee commitments and NRC requirements, including the requirements of the applicable cod The code applicable to the material and work addressed by the inspector is ASME Section XI (74S75).
9 Review of Procedures (74S75)
(1)
The following !SI procedures were reviewed for proper approval and for personnel qualification requirements:
Title Preservice and Inservice Inspection of Reactor Vessel Remote Visual Examination of Reactor
- Vessel Internals Manual Ultrasonic Examination of Full Penetration Circumferential Welds and Long1tudinal Butt Welds Liquid Penetrant Examination Procedure Visual Examination Procedure Procedure N ISI-154, Rev. 3 ISI-88, Rev. 1 ISI-205, Rev. 2, Amendment 3 ISI-11, Rev. 9, Amendment 3 ISI-8, Rev. 8, Amendment 1, Field Change 3/83 (2)
The NRC inspector questioned the licensee's corporate Level III examiner as to what kind of transducer was being used to ultra-sonically examine Surry 1s cast stainless steel pipin The examiner informed the inspector that their contractor, Westing-house, would perform the examinations using 41° longitudinal wave water column transduce The inspector noted that the transducer referred to had been found lacking in examination of some cast stainless steel piping - as documented in NRC Inspection Report 395/81-22; and that more recently developed transducers appeared to provide much better examination The adequacy of the trans-ducer to be used in examination of Surry 1s cast stainless steel piping was identified by the inspector to the licensee as Unreso 1 ved Item 280, 281/85-15-02, Adequacy of Transducer for Examination of Cast Stainless Steel Pipin The inspector indi-cated that this matter would remain unresolved pending the licen-see's demonstration of the transducer's adequacy in a subsequent NRC inspectio *observation of Work (737538)
(1) Qualifications and Certifications of Examiners The inspector reviewed qualifications and certifications of one of each level of examiners to verify that they properly reflect: *
(a)
Employer's name (b)
Person certified
(c) Activity qualified to perform (d)
Currently qualified (e)
Signature, title, and level of certifying individual (f) Annual visual acuity and color vision examination (2) Observation of Automated Ultrasonic Examination of Reactor Vessel The inspector observed a portion of the Field System Calibratio In addition, the inspector observed and discussed with the Level III examiner video tape and computer printout data for the following: *
outlet nozzle weld 10 printout data outlet nozzle weld 14 video tape data girth weld 2 printout data for indications #4 and #47 The above were observed and discussed with the Level III examiner to verify the following:
(a)
Procedure available and being followed (b)
Personnel knowledgeable of examination method and equipment operation (c) Personnel properly qualified for functions being performed (d)
Proper recording of data (e)
Proper apparatus in use (f) Adequate coverage (g)
Calibration in accordance with requirements (h)
Proper sizes and frequencies of search units (i) Proper recording of indications (j) Accurate orientation of reference points Review of Information Related to Technical Spectfication Changes (92706B) -
Units 1 and 2 The inspector reviewed and discussed with cognizant licensee personnel, information related to Technical Specification (TS) changes requested 'by the licensee in a letter to the NRC dated September 21, 198 The proposed changes are currently under evaluatio The review and discussions were undertaken by the inspector to better understand the proposed changes, their bases, their effect on previous commitments and their impact on inspections to be conducted by NRC Region I As a consequence of the review and discussions, the licensee discovered that certain ISI reporting requirements had been removed in 1976 in Amendment No. 14 to the TS Also as a conse-quence of the review and discussions, the licensee agreed with the inspector that they would need to provide technical justification for deletion of augmented ISis, as proposed by the TS chang They stated that they now tentatively plan three licensing related submittals as a consequence of the review and discussion:
~
.,
- ,I,." t
!'i' *~
(1)
Modification to the original TS change proposal of September 21, 1982, omitting the changes to augmented ISI requirement Note:
It was considered that evaluation and approval of this change could be more quickly obtained as less extensive technical justifica-tion would be require (2)
Submittal of a proposed TS change to delete certain augmented ISI, accompanied by detailed technical justification (3)
A submittal to address the reporting requirements omitted in 197 The above are essentially licensing issues and were not individually ident-ified for inspection followu However, the augmented ISI requirements themselves and their bases were identified by the inspector for further review in subsequent NRC inspection, as Inspector followup item 280, 281/85-15-03, Verification of Augmented Inservice Inspection.
Inspector Followup Items (IFis) (927018)
(Closed) IFI (280, 281/83-19-02):
Evaluation of Reflectors in Steam Genera-tor Upper Shell to Transition Cone Wel The licensee has completed evaluation of the reflectors with examinations described in detail in paragraph 5 above.