IR 05000244/1993023
| ML17263A502 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/09/1993 |
| From: | Modes M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17263A501 | List: |
| References | |
| 50-244-93-23, NUDOCS 9312270228 | |
| Download: ML17263A502 (9) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
REPORT/DOCKET NO.
50-244/93-23 LICENSE NO.
LICENSEE:
FACILITYNAME:
INSPECTION AT:
DPR-18 Rochester Gas and Electric Corporation Ginna Nuclear Power Station Rochester, New York INSPECTION DATES:
November 15-19, 1993, INSPECTOR:
Joseph E. Carrasco, Reactor Engineer Materials Section, EB, DRS Date APPROVED BY'ichael C. Modes, hief Materials Section, EB, DRS
/z eq9 Date 9312270228 931216 PDR ADOCK 05000244
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Pl determine whether the licensee's design and modification of safety-related piping and pipe supports are performed in accordance with regulatory requirements, engineering specifications and properly documented.
This inspection included a review of the current status, organization, design features, and project schedule for the steam generator replacement project.
This inspection also included a review of the steam generator "B" hydraulic snubber's operability evaluation.
During the course of the inspection, the November 17 plant shutdown for the steam leak repair was closely monitored.
~exult:
Based on the limited sample reviewed, it was concluded that the analysis of the reactor heat removal, RHR-100, segment of piping and associated supports was found to be satisfactory and in compliance with the licensee's established Design Criteria No. EWR-2512.
It was also concluded that Procedure MDG-21 satisfactorily addresses the NRC concerns from a previous piping inspection.
The implementation of MDG-21 resulted in a computerized database which offers an adequate means to maintain control, status and design information iri a user friendly manner.
It was found that the licensee actions to address the loose vent plug on snubber SGB-4 were acceptable.
Although the licensee has taken the proper steps to correct and to prevent future leakage of the snubbers, the inspector emphasized the importance of closely analyzing this supporting arrangement (two hydraulic snubbers and six structural bumpers) for the new proposed steam generators.
The licensee's steam generator replacement project (SGRP) is in-place and functional.
The design features for the new steam generator are well defined, the proposed cutouts in the containment dome are receiving adequate analytical attention, and the SGRP's schedule is well defined.
At the time of the exit meeting, the inspector did not receive supporting documentation for the root cause for the leak of blowdown line S/G "B". A clear explanation of why the blowdown line of S/G "A" was replaced under an erosion corrosion program was not available.
This lack of formal documentation constitutes an unresolved item (URI 244/93023-01).
1.0 PURPOSE AND SCOPE DETAILS A safety inspection was conducted at the Ginna Nuclear Power Plant to determine whether the licensee's design, and modification of safety-related piping and pipe supports were performed in accordance with regulatory requirements, engineering specifications and were properly documented.
This inspection included a review of the current status, organization, design features, and project schedule for the steam generator replacement project.
This inspection also included a review of the steam generator "B" hydraulic snubber's operability evaluation.
During the course of the inspection, the November 17th plant shutdown for the steam leak repair was closely monitored.
2.0 REVIEW OF THE R. E. GINNA SAFETY-RELATED PIPING AND PIPE SUPPORTS (37700)
~B'~ck roun A review of the Ginna Station s safety-related piping and pipe support design bases and associated calculations was performed.
The NRC issued industry noticesBulletin 79-02 dealing with concrete expansion anchors used in pipe support base plates, and 79-14 which dealt with seismic analysis of as-built safety-related piping.
In addition to the activities conducted to satisfy the requirements of Bulletins 79-02 and 79-14, Rochester Gas and Electric Corp. initiated a systematic evaluation program (SEP).
The inspector verified that the SEP topics affecting the safety-related piping and pipe supports were approved by the NRC via a letter to the licensee, dated November 15, 1989.
These topics are (SEP topic III-6) seismic design considerations and (SEP topic III-7B)load combinations.
Based on these approved topics, the seismic upgrade program for Ginna was initiated under engineering work request (EWR) 2512.
This program included the generation of new floor response spectra, new spectral shapes and damping values, and a complete re-analysis of the safety-related piping and associated supports.
Piniong Using the seismic upgrade program data base, the inspector randomly selected stress analysis
"RHR-100," which corresponds to a segment of the residual heat removal system (RHR).
The selected piping segment extends, from the containment penetration No. 111 to the reactor vesse The inspector verified that the design parameters documented in the FSAR were properly input in stress analysis RHR-100.
The verified design parameters are the piping system's temperature, pressure, materials, schedule and sizes.
The inspector also verified that the geometric configuration of the computer model used in the analysis matched the as-built isometric shown on drawing C-381-354, sheet 3, Revision 4. It was also verified that the correct allowable stresses were used and the result of the analysis was within the allowable stress limitof the code of record.
The inspector selected two supports for review from the stress isometric:
RHU-5 and RHU-8. For these supports, the inspector verified that the resulting loads at the corresponding nodal points of the pipe stress analysis were properly applied in the support calculation to qualify these supports.
The inspector verified that the support's components were within the American Institute of Steel Construction (AISC) code allowable and the welds were sufficient for the given qualifying loads.
Qgnclu~in Based on the sample reviewed, the inspector concluded that the analysis of the RHR-100 segment of piping and associated supports were satisfactory and in compliance with established Design Criteria No. EWR-2512.
The inspector had no further questions regarding the adequacy of the seismic upgraded piping and associated supports.
3.0 REVIEW OF PIPING SEISMIC UPGRADE PROGRAM'S DATABASEAND DOCUMI~22TATION (37700)
In response to a concern regarding the control and documentation of the seismic upgraded piping and supports, the licensee initiated a program to control and maintain the seismic upgrade program in a computerized system.
This seismic upgrade program is governed by a Mechanical Design Guide (MDG) No. 21.
The inspector noted that MDG-21 has a well established scope, defining the piping systems affected by the seismic upgrade program (SUP).
In this design guidance, the piping scope was defined in terms of the analyzed piping stress problems and their corresponding isometric drawings.
To assess the licensee's database the inspector was given a practical demonstration.
In this demonstration, the inspector noted that the seismically upgraded piping and supports have computerized retrievable drawings in the form of isometrics, piping layouts and each stress line has its corresponding pipe supports drawings, which are also retrievable.
The inspector also verified that any changes in a particular drawing resulting from a pipe modification are properly controlled with prescribed forms.
These forms are reviewed, authorized and concurred by the cognizant manager or the lead mechanical engineer before the data is input in the databas g~nl i~in The inspector concluded that Procedure MDG-21 is adequate and satisfactorily addressed the NRC concerns from previous inspections:
The implementation of MDG-21 resulted in a computerized database that is a useful tool and is an adequate means to maintain control, status and design information.
4.0 REVIE% OF THE STEAM GF22ERATOR "B" SNUBBER HYDRAULICLEAK Qg~kr i~n During a routine preventive inspection (PM) performed by the licensee on 9/29/93, hydraulic fluid was found on the secondary level platform around the steam generator (S/G) "B" and the common hydraulic fluid reservoir for the S/G's snubbers SGB-3 and SGB-4 was found empty.
Subsequently, a containment entry was made by the licensee to inspect the snubbers and the attached hydraulic tubing, to determine the source of the leakage.
During this inspection, hydraulic fluid was found leaking out of a loose vent plug on snubber SGB-4.
The common reservoir was filled with hydraulic fluid and the loose vent plug was tightened along with.the fittings on the tubing.
Subsequent inspections have shown that the hydraulic fluid level in the reservoir has remained steady.
~Findin s The inspector interviewed the licensee's mechanical engineer responsible for the preparation of Design Analysis DA-ME-93-127, Revision 2, which was prepared to evaluate the effect that the loss of Quid of snubbers SGB-3 and SBG-4 had on the integrity of the reactor coolant system (RCS).
The intended function of snubbers is to allow displacement due to deadweight and thermal loadings and to restrict displacements caused by seismic events, water hammers, and other transient forces.
Because the degraded condition of these snubbers involved the loss of hydraulic fluid, the ability of the snubbers to move in response to deadweight and thermal loads was not impaired.
However, the loss of fluid experienced by snubbers SGB-3 and SGB-4 precludes the snubbers ability to lock in the event of a seismic event.
Therefore, the licensee's evaluation concluded that the structural integrity of the RCS had not been adversely impacted by the loss of fluid experienced by snubbers SBG-3 and SBG-The inspector found the licensee's design analysis to be logical and acceptable.
However, the inspector requested a root cause analysis and preventive measures to preclude the repetition of this situation in the future.
The licensee presented the work order (WO) No.
9341525.
In this WO, the licensee included an existing failure cause analysis and corrective action.
This document stated that the last quarter snubber inspection (M-40) completed 7/8/93, found all snubbers to be in good condition.
This prompted the licensee to conclude that the leak in snubber SGB-4 was caused by a vibration or a human error during inspection and cannot be associated with a generic condition.
As a corrective action, the licensee has initiated a procedure change in Procedure M-40.7,
"Steam Generator Snubber Inspection and Maintenance," to incorporate a step for tightening the vent plug and tubing fittings after completion of maintenance and inspection work.
g~nlu~in The inspector found the licensee's actions to address the loose vent plug on snubber SGB-4 acceptable.
The inspector had no further questions in this regard.
5.0 STEAM GENERATOR REPLACEMENT PROJECT (SGRP)
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generators during the Spring 1996 refueling outage.
The decision to proceed with replacement was based upon the projected trends in the steam generator tube defect population and associated tube plugging and sleeving, in secondary side heat transfer fouling, and the resulting loss of heat transfer capacity caused by these trends.
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established an in-house project team reporting directly to the Vice President-Nuclear Production.
It was noted that the licensee's intent is to internally manage and technically effect the design, fabrication, and installation of the replacement steam generators.
The licensee has established an open, cooperative relationship with both the selected designer/fabricator and the selected installer to ensure the ready availability of critical design information within the Project.
In addition, the licensee continued to work with the original nuclear steam system supplier designer to ensure that the system design bases is preserved.
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ld have increased margins (20%) against plugging and secondary heat transfer fouling.
The replacement steam generators would use the latest industry technology to minimize the potential for tube defects and plugging and to improve maintenance.
Specifically, these design improvements include:
Inconel 690 tubing Lattice Bar Support Grids (410 Stainless)
Enhanced U-Bend Supports (410 Stainless)
Full-Depth Hydraulic Tube-end Expansion 18" Primary and Secondary Manways Inspection Ports at Each Support Grid Elevation The licensee added that an enhanced feedwater distribution system would be installed to minimize the potential for waterhammer.
In addition, a flow-limitingventuri would be installed in the steam outlet nozzle'o mitigate the containment pressurization effects of a postulated main steam line break.
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SGRP showing the replacement is based on a one-piece replacement steam generator design installed via cutouts in the containment dome.
Design and licensing analysis associated with the replacement steam generators would be completed in early 1995.
The steam generators would be delivered in February 1996.
Design and licensing efforts associated with installation, including analysis of the containment cutout scheme willbe completed in mid-1995.
Construction of support facilities on-site will take place during 1995 with the actual replacement scheduled for a Spring 1996 outage.~ The licensee informed the inspector that the majority of major parts and components (forgings, shell plates, tubing) are currently being fabricated by the subvendors.
Vessel stress, thermal-hydraulic, and licensing support analysis are in-progress.
Generator fabrication at the vendor's facility began in early 1993.
The project staff has identified that initial walkdowns of the containment were made during the Spring 1993 refueling outage.
Formal design interfaces were established between the licensee and the contractor.
Detailed design work required for the installation effort is scheduled to begin in early 1994.
The licensee has established a separate effort to plan and manage the radiation protection efforts associated with steam generator replacement.
Qgnclu~in The SGRP's organization is in-place and functional, the design features for the new steam generator are well defined, the proposed cutouts in the containment dome are receiving adequate analytical attention, and the SGRP's schedule is well define.0 PLANT SHUmOWN FOR STEAM LEAKREPAIR Baack round During the course of this inspection, the inspector was informed that on November 17, 1993, the licensee conducted a controlled shutdown to repair a secondary plant steam leak in containment.
Containment atmosphere radiation monitors indicated that the leakage was not RCS coolant.
Subsequently, it was determined that the source of the leak was found to be the "B" steam generator blowdown line. The leakage rate, as determined using the containment water inventory monitoring system, was approximately 300 gallons per day. At this point, the licensee concluded that shutdown was necessary to support repairs because the at-power general area radiation field was 11 rem per hour, and because the leak cannot be isolated from the steam generator.
~in i~in The licensee was going to seek approval under GL 90-05 to repair outside the code.
The repair was encapsulation of the elbow and affected segment of the blowdown line steam generator (S/G) "B." However, the licensee decided to cut the pipe and to replace the affected segment of piping and the elbow, including valve 5706 which had a history of leaks with a like-for-like segment.
Since the newly proposed repair involved a code allowed like-for-like repair, a GL 90-05 submittal was not required.
Therefore, the leak repair took place, followed by the pertinent post-modification test to assure the adequacy of the repair.
Qgnclu~in At the time of the exit meeting, supporting documentation for the root cause of the leak of the blowdown line for S/G "B" was not available.
Further, the basis for why the blowdown line of S/G "A" was replaced under an erosion corrosion program was not available.
This lack of formal documentation is an unresolved item pending the NRC review of the supporting explanations and root cause analysis (URI 244/93023-01).
7.0 MANAGElVEÃTMEETINGS Licensee management was informed of the scope and purpose of the inspection at the beginning of the inspection.
The findings of the inspection were discussed with the licensee management at the November 19, 1993, exit meeting.
The licensee did not express any objections to the findings of this inspection during the exit meeting.
See Attachment 1 for attendanc ATTACHMENT1 Persons Contacted h
r EI E. K. Voci J. F. Smith D. R. Markowski D. Morgan L. Rochino G. Wrobel B. J. Carrick R. Jaquin W. Tono F. Gilbert A. Borodotsky M. Lilley K. Muller Manager, Mechanical Engineering Manager, Steam Generator Replacement Project (SGRP)
Lead Mechanical Engineer Lead Mechanical Engineer Lead Mechanical Engineer Manager, Nuclear Safety Lead Engineer (SGRP)
Nuclear Safety Engineer Mechanical Engineer Mechanical Engineer Erosion Corrosion, Engineer Coordinator Director, SGRP's QA IST Engineer 1 rR I t mmi i n T. Moslak Senior Resident Inspector denotes those piesent at the exit meeting