IR 05000324/1993034
| ML20057B501 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/08/1993 |
| From: | Blake J, Coley J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20057B492 | List: |
| References | |
| 50-324-93-34, 50-325-93-34, NUDOCS 9309220145 | |
| Download: ML20057B501 (9) | |
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UNITED STATES
'yg [Se nog .g*e NUCLEAR REGULATORY COMMISSION REGloN H ; *E 101 MARIETTA STREET, N.W., SUITE 2900 i '$ s ATLANTA, GEORGIA 30323 0199 S %,, ....+
f Report Nos.: 50-325/93-34 and 50-324/93-34 Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 Docket Nos.: 50-325 and 50-324 License Nos.: DPR-71 and DPR-62 facility Name: Brunswick 1 and 2 Inspection Conducted: August 16-20, 1993 Inspector: \ . ) ,o tu ?-7-73 Date Signed
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J:CLMoley Approved by: [[ht49 J. J. Blake, Chief ( L- dvv E b Date Signed
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MaterialsandProcessesSe/ ction ' Engineering Branch Division of Reactor Safety SUMMARY ; Scope: i l This routine, announced inspection was conducted in the areas of inservice I inspection, review of in-vessel visual examinations'for the Unit I reactor core support shroud and core spray sparger piping, review of Unit 1 ! radiographic film packages, review of and expansion of scope for a material traceability concern initially addressed by the . site authorized nuclear inspector (ANI), and observation of work activities involved in obtaining boat sample specimens of cracks on the internal and external surfaces of the reactor vessel shrou Results: In the areas inspected, violations or deviations were not identifie As a result of ISI data reviewed and discussions held with cognizant management personnel, the inspector concluded that the licensee Ms made significant improvements in their structured program to manage intervice inspection (ISI) information and work activitie ;
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930922014S 930914 PDR G ADOCK 05000324 PDR [D[5; E-e ' l
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One Unresolved Item was identified. This item concerned the need for the licensee to conduct an audit to determine if failure of plant personnel to follow procedures for uniquely identifying materials issued from the storeroom has resulted in failure to maintain traceability of replacement items to , specific safety related components (paragraph 4).
Two additional weaknesses were addressed to CP&L Managemen One dealt with failure of a nondestructive test examiner to be attentive to detail when ' calibrating a radiographic densitometer (paragraph 3). The second weakness dealt with examiner acceptance of very minimal radiographic quality for radiograph film of thin wall (.237") piping welds when corrective measures are available to improve the quality (paragraph 3).
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REPORT DETAILS 1.- Persons Contacted , l Licensee Employees l W. Allen, Supervisor, Material Control
*G. Barnes, Manager, Operations, Unit 1 l
*H. Bordeaux, Manager, Quality Control
*M. Brown, Plant Manager, Unit 1
*J. Crider, Manager, Inservice Inspection
*J. Dobbs, Special Assistant to Site Vice President
*L. Gard, Manager, Engineering Projects
*S. Hardy, Senior Engineer, Nuclear Engineering Department
*C. Hinnant, Director of Site Operations i
*T. Jones, Senior Specialist, Regulatory Affairs
*R. Lamb, Project Engineer
*J. Langdon, NDE Supervisor
* Levis, Manager, Regulatory Affairs (Acting)
* Miller, Manager, Technical Support
*R. Sims, Senior Engineer, Nuclear Engineering Department Other licensee employees contacted during this inspection included engineers, technicians, and administrative personne NRC Resident Inspectors
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*P. Byron, Resident Inspector
*R. Prevatte, Senior Resident Inspector
* Attended exit interview Acronyms and initialisms used throughout this report are listed in the last paragrap . Inservice Inspection - Review of In-vessel Visual Examination Data !
For the Unit 1 Reactor Vessel Shroud and Core Spray Sparger Piping j This inspection was conducted to review the special processes used by the licensee and their vendor (General Electric) during the in-vessel visual inspection of the reactor core support shroud and the core spray i sparger piping. The core support shroud is the reactor internal component which surrounds the core and directs coolant flow. The visual examinations on the shroud top guide support ring identified a crack which was assumed to extend 360 Degrees F* around the shroud support : ring. The crack was adjacent to the toe of circumferential weld H-3 and i on the core side of the support ring. In addition, a one inch axial i crack adjacent to the lower side of circumferential weld H4 was identified on the outside surface of the shroud. The in-vessel visual examinations also had -identified a 3" long crack in the B-loop of the core spray sparger at the upper tee box circumferential weld heat affected zone, and a 4" long crack on the lower sparger arm 18 inches from the tee box in the t-
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l 2 l circumferential weld heat affected zone. The function of the internal core spray sparger is to provide a contained flow path to direct water to the core region in the event of a LOC Summary of The Core Support Shroud Activities Reviewed The visual examinations on the shroud had been performed by the I licensee as recommended in GE's Rapid Information Communication Service Information Letter (RICSIL) No. 054. In order to observe the examinations that had been performed, the inspector reviewed video tapes of the applicable in-vessel visual inspections on the shroud. The visual examination was performed by lowering an under water camera into selected top guide cells around the periphery of the core. Since the entire surface of the shroud was not accessible, the length of the crack could not be determine However, the crack indication appeared to be continuous. The i following video tapes were reviewed by the inspector to determine the severity of the indications: Video Taoe N Weld and Cell N BNP1 - 93-13 H-3, (ID) Cell 10-47 BNP1 - 93-13 Vertical weld, (10) Cell 06-11 BNPI - 93-13 H-3, (ID) Cell 06-11 BNP1 - 93-13 H-3, (ID) Cell 10-47 BNPI - 93-13 H-3, (ID) Cell 46-43 BNP1 - 93-13 H-3, (ID) Cell 42-07 BNP1 - 93-14 H-3, (ID) Cell 50-19 t BNP1 - 93-25 H-4, (0D) @ 315' Axial indication BNP1 - 93-25 H-3, (ID) Retake From Cell 46-43 . BNP1 - 93-17 H-4, (ID) 0 315' to Determine if the ! Indication Penetrated Vessel Wall In addition to the visual examinations, the licensee attempted to ultrasonically size the indications using a transducer attached to i a fixture on a pole. This technique had been qualified by the licensee on a plywood mockup of the applicable area of the shrou Video's were not taken of the ultrasonic signals displayed on the instrument's screen when the sizing examinations were performe However, a video was taken of the transducer placement and of the limitations encountered. Based on review of the video tape and observing the limitations encountered by the transducer, the l inspector was surprised that the licensee had been able to detect j crack tip signals, The maximum indication depths obtained by this i technique was 0.4 inches for the support ring indication (above weld H-3) and 0.25 inches for the axial indication at weld H- Transducer positioning for the following fuel cell locations were l reviewed by the inspector: Surface Examined Weld N Cell N Ecan Anale Outside H-3 06-11 60' RL Outside H-3 42-07 60* RL _ _ . -. _
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l l Fuel Cell locations Continued l Surface Examined Weld N Cell N Scan Anale Outside H-3 46-43 60* RL " Outside H-3 50-27 60* RL Outside H-3 50-19 60* RL Outside H-3 10-47 60* RL Outside H-3 06-11 60* RL Inside H-3 06-11 70* RL Inside H-3 10-47 70* RL Inside H-3 46-43 70* RL -
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Inside H-3 42-07 70* RL Inside H-3 50-27 70* RL Inside H-3 50-19 70* RL Inside H-3 46-43 70* RL Outside (Axial Ind.) H-4 315' 70* RL At the time of this inspection, the cause of cracking had not been identified. However, boat samples of cracks in both welds H-3 and H-4 were removed for metallurgical analysis. Paragraph 5 describes the process used by the licensee to remove the boat sample The process used for sample removal was observed by the inspector. The boat samples will be analyzed by GE cod appropriate measures taken by the licensee based on the metallographic result ' b. Summary of Core Spray Sparger Piping Activities Reviewed - Unit 1 In accordance with NRC Bulletin 80-13, the reactor pressure vessel , internal piping and spargers associated with the core spray system , are visually examined with a remote operated underwater camera , during each refueling outage. Although, cracking in the in-vessel core spray spargers and piping is an industry concern, prior to this outage, no crack indications had been reported from Unit I examinations. However, Unit 2 is currently operating with two cracks in the reactor pressure vessel internal core spray piping and spargers. One crack is in the 90 Degrees F* azimuth (north loop) piping-to-tee box circumferential weld and the other is in the upper sparger tee-to-arm circumferential weld. These flawed pipe /spargers have been repaired by the using brackets / clamps which provide full structural reinforcement to the piping, equivalent to a welded join During the present Unit 1 Refueling Outage (No. 8), two linear indications were found by visual examinations using a remote ' , operated underwater camera on the B-loop core spray piping. One indication is approximately 4" long and is in the heat affected . zone of a circumferential weld located in the in-vessel piping , between the B-loop inlet nozzle and the sparger. This weld is , located approximate 18" downstream of the B-loop tee bo .
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The other indication was found on a tee-to-sparger arm circumferential weld on one of the lower B-loop spargers. This indication is approximately 3" long. The tee and sparger are located inside the shroud at a lower elevation than the inlet - piping containing the other indication. The inspector reviewed ! video tapes BNPI-93-4 and BNP1-93-5 of the core spray in-vessel ; visual inspection in order to verify the licensee's examination ' effort The indications were evaluated by General Electric for: (1) the effect on the structural integrity of the in-vessel core spray piping, (2) the effect of leakage through the assumed through wall indications and the impact to the ECCS analysis, and (3) the effects of any postulated loose parts on safety related equipment in the pressure vessel or the effect on in-vessel components. The conclusion of the analysis was that the Unit 1 internal core spray . piping is acceptable in the as-found condition for the next ' operating cycl Within the areas examined, no violation or deviation was identifie . 3. Review of Unit 1 Radiographic Film and Associated Records The inspector reviewed the radiographs listed below to determine whether ] thef were prepared, evaluated, and maintained in accordance with CP&L's 1 approved radiographic procedure and Sections III and V of the ASME Cod Radiographs in each film package were also reviewed to determine whether i the following examination parameters had been correctly adhered to: film l quality; penetrameter type, size, placement, and sensitivity; film ! density and density variation; film identification; and weld coverag Radiographs for the following ASME Class 2 carbon steel pipe welds were reviewed: ,
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Weld Identification Pine Thickness and Size 1-ELL-481 .237" X 4" Diameter 1-ELL-460 .375" X 16" Diameter 1-ELL-439 .375" X 16" Diameter 1-ELL-441 .375" X 16" Diameter l l-ELL-337 .237" X 4" Diameter { 1-ELL-433 .375" X 16" Diameter ' l-ELL-434 .375" X 16" Diameter 1-ELL-336 .375" X 16" Diameter . 1-ELL-477 .375" X 16" Diameter l l-ELL-338 .237" X 4" Diameter During the review of the above radiographs, the inspector noted two weaknesses, which, if allowed to continue, could adversely affect the radiographic process. The first weakness dealt with the calibration of the radiographic densitometer. When the inspector checked the densitometer calibration prior to reading the above radiographs, the
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instrument (S/N 8732) e round to be saturated at a density reading of 3.8 However, the calibration log indicated that the densitometer had been successfully calibrated to a density of 3.9 earlier that mornin i The inspector and the CP&L Level III examiner's initial attempts to re- l calibrate the instrument were unsuccessful. The Level III examiner subsequently found that the reason why the densitometer could not be l calibrated was because of dust on the inside surface of the densitometer causing the high intensity light to be refracted. The Level III re-examined previously accepted film and determined that the discrepancy had not effected the evaluation of the film. Since no film had been read in over 30 days, it appears that the discrepancy had incurred during the period of inactivity. To prevent the discrepancy from recurring the licensee informed the inspector '. bat a preventive
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maintenance procedure would be written and performed on a periodic basis. In addition, personnel performing the densitometer calibration
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will be instructed to be more attentive to detail when calibrating equipment used to accept or reject ASME Code radiograph The other weakness observed by the inspector concerned film quality level. During the review of radiographs for thin wall piping (.237" wall thickness), the quality level for the contact technique using T film was very minima Improvement in quality was needed, especially when radiographing existing pitted piping. The inspector pointed out that improvements can be accomplished by using a higher R factor film or increasing the source to film distance. The licensee plans to experiment with the exposure variables available and improved the radiographic qualit Within the areas examined, no violation or deviation was identifie . Material Traceability Concern The inspector reviewed an Engineering Work Request (EWR) Submittal (N ) and held discussions with the site Authorized Nuclear Inspector, who had initiated the EWR, to determine what effect this reported item had on installed materials in both Units 1 and 2. In the EWR, the ANI had requested the licensee to evaluate the requirement for traceability to unique Certified Mill Test Report (CMTR) for individual replacement items within ASME Section XI Class I, 2, and 3 boundaries. The ANI also stated that CP&L currently was using the purchase order number as the unique identifier and that a purchase order may have several CMTR's'for a specific item. The ANI had addressed this concern because a unique identifier was needed in order for him to complete the NIS-2 Reports, which are submitted to NRC with the NIS-1 Reports. Attached to EWR N was a completed NIS-2 Form to be used as an example that purchase orders were used as the unique identifie After reading EWR No. 12684 and discussing the matter with the ANI, the inspector's immediate concern was whether ASME Code replacement items were traceable to the specific material certifications that supported the item. The inspector requested a meeting with cognizant CP&L personnel to discuss this matter and requested that the licensee '
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demonstrate that materials issued from storage are uniquely identifie The sample used during the demonstration was the sample NIS-2 form which the ANI's EWR had provided. The audit of material requisition sheets for the sample NIS-2 form disclosed that two items had been correctly identified with a purchase order number and a unique heat numbe l However, five separate items were issued on material requisition forms, I which had not been completed in accordance with CP&L's Procurement and Material Control Procedure No. 0-PMC-04, Rev. 27, in that, the fastener heat numbers had not been documented as required. The licensee immediately issued Adverse Condition Report No. 93-281 for failure of personnel to follow procedure when filling out the requisition issue forms. Further review revealed that for the purchase orders involved in the five items found discrepant, only one CMTR was applicable to each of the purchase orders. Therefore, traceability for these five items had not been los However, the above audit revealed that material issue room attendants .
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are not following procedure in uniquely identifying materials they issue. What affect this discrepancy has had on the traceability and control of safety related replacement parts will have to be determined by the licensee. This issue will be tracked as Unresolved Item No. 50-325,324/93-34-01, Traceability of Safety Related Replacement Part During the inspector's discussion of this item at the exit meeting, senior plant management stated that the ACR issued by CP&L on this discrepancy had been elevated to severity level 1 item and that a complete investigation would be performed by the license Within the areas examined, no violation or deviation was identifie . Removal of Crack Samples from the ID and OD Surfaces of the Shroud The inspector observed portions of the work activities performed by GE and the licensee to obtain a boat sample from the circumferential crack above weld H-3 on the shroud top guide support ring. The sample was being retrieved for metallographic evaluation in order to determine the cracking mechanism. The sample observed by the inspector was being removed at fuel cell No. 42-07. The cutting tool used to retrieve the sample was installed from the refueling bridge by lowering it into place with handling poles. Two underwater cameras were also lowered into position through the fuel cells. The cutting tool was connected by cables and hoses to the electrical discharge machining (EDM) equipment which was used to cut the samples from the shroud. Positioning the fixture holding the cutting tool and maintaining camera surveillance proved to be a timely challenge. Once the equipment was positioned, the cutting process was equally slow. However, the cutting was successfully performed in accordance with CP&L's Special Procedure No. ISP-93-05 Personnel involved in the retrieval operation (CP&L and GE) were well versed in their responsibilities and carried them out in an impressive manne A sample was also retrieved from the axial crack on the shroud OD at Weld H-4. This cut was not observed by the inspecto .,
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Within the areas examined, no violation or deviation was identifie . Exit Interview The inspection scope and results were summarized on August 20. 1993, with those person indicated in paragraph I. The inspector described the areas inspected and discussed in detail the inspection results listed , below. Proprietary information is not contained in this repor , Dissenting comments were not received from the license ,
(0 pen) Unresolved Item No. 50-325,324/93-34-01, Traceability of Safety Related Replacement Parts, paragraph . Acronyms and Initialisms !
ACR - Adverse Condition Report ANI - Authorized Nuclear Inspector ASME - American Society of Mechanical Engineers . BNP - Brunswick Nuclear Project CMTR - Certified Mill Test Report CP&L - Carolina Power and Light ECCS - Emergency Core Cooling Systems EDM - Electrical Discharge Machining EWR - Engineering Work Request , GE - General Electric ; ID - Inside Diameter ' IND - Indication ISI - Inservice Inspection LOCA - Loss Of Coolant Accident N Number NDE - Nondestructive Examination NRC - Nuclear Regulatory Commission j OD - Outside Diameter i RICSIL- Rapid Information Communication Service Information Letter l RL - Refracted Longitudinal l S/N - Serial Number l l I i
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