IR 05000244/1994015
| ML17263A735 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 07/08/1994 |
| From: | Kay L, Eugene Kelly, Prividy L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17263A734 | List: |
| References | |
| 50-244-94-15, NUDOCS 9407200153 | |
| Download: ML17263A735 (16) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
DOCKET/REPORT NO.
50-244/94-15 LICENSEE:
Rochester Gas and Electric (RG&E) Corporation Rochester, New York 14649 FACILITY:
DATES:
Robert E. Ginna Nuclear Power Plant June 6-10, 1994 INSPECTORS:
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Leonard Prividy, r. Reactor Engineer
~ Systems Section Division of Reactor Safety Date ne ~, Reactor Engineer Electrical tion Division of Reactor Safety Date )'-
APPROVED BY:
Eug eK 1, Chief Systems ion Division Reactor Safety Da
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plant modifications were being conducted in accordance with controlled procedures, and in accordance with NRC requirements.
Other engineering activities were also reviewed to assess the effectiveness of the engineering organization in providing technical support for safe plant operation.
gmi~lt: The inspector concluded that the design change and modification program was being controlled well, with clearly detailed administrative procedures.
Several modification packages reviewed were well engineered.
Information was not yet available regarding several inspectors'omments on main feed regulating valve changes that resulted in an unresolved item.
The engineering organization has initiated efforts to prioritize and track items in the engineering work backlog, to further develop design basis documents, and to develop a configuration management information system.
9407200i53 940708 PDR ADOCK 05000244
1.0 INSPECTION SCOPE DETAILS A well-implemented design change and modification program assures that changes to the plant do not degrade safety systems, structures, and components.
Effective engineering and technical support in resolving problems also ensures safe plant operation.
The objective of this inspection was to verify that changes and repairs to plant components and systems, which are described in the final safety analysis report (FSAR), were implemented per controlled administrative procedures that satisfy regulatory requirements.
This objective was accomplished by reviewing several modifications and engineering work items to evaluate engineering involvement and problem resolutions.
Other new and ongoing corporate engineering programs were also assessed from a safety and performance perspective.
2.0 INSPECTION FINDINGS 2.1 Design Changes and Modifications (37550)
To ascertain that design changes and plant modifications were performed in accordance the requirements of the Code of Federal Regulations and plant licensing documents, the inspectors reviewed the current procedures and selected modification packages planned for the upcoming 1995 outage, or implemented during the past 1994 outage.
This review evaluated the technical quality of the modifications, the thoroughness of the design analysis, design input, technical reviews and safety evaluations, and management involvement in the resolution of problems.
The inspectors reviewed the process for preparation, review, and approval of 10 CFR 50.59 safety evaluations to verify proper evaluation of plant design basis when performing modifications.
This process was explained in Engineering Procedure EP-3-P-135, Revision 0, "Preparation, Review, and Approval of 10 CFR 50.59 Safety Evaluations."
The RG&E design modification process is described in Nuclear Engineering Quality Standard NEQS-2, Revision 0, "Nuclear Engineering Services Modification Design Process Standard."
This standard establishes a uniform method for performing engineering work requests (EWRs) at Ginna. It includes a description of the interrelationships of those procedures required to develop engineering information for modifications.
The process by which safety evaluations ensure that proposed plant changes do not compromise plant safety and satisfy NRC requirements was included in Engineering Procedure QE311, Revision 6, "Preparation, Review And Approval Of Safety Analyses."
Ginna utilizes many engineering procedures to document reviews and verifications made throughout the modification process.
These reviews and verifications make up independently approved documents, including a conceptual design document by the system engineer, a
design criteria document, a fire protection review, a design verification by an independent
and qualified reviewer, an integrated design review completed by all affected departments, an installation specification, and pre-operational and post-installation test specifications to confirm expected results.
The coalition of these documents complete the modification process throughout the following phases of development:
design input and preliminary design development; detailed design; installation, test, and turnover; as-builting; and administrative close-out.
2.1.1 Undervoltage Protection System Design - EWR No. 5282C EWR No. 5282 C is a modification completed during the last refueling outage to provide undervoltage protection to mitigate a loss of voltage to 480 Vac safeguard buses.
This modification changed the power source from a 120 Vac supply to a 12 Vdc supply to provide continuous power to the 480V safeguard buses.
A review was made of the completed modification process including the Conceptual Design, Design Criteria, Appendix R Conformance Review, Design Verification, Integrated Design Process, Installation Specification, Pre-Operational Test Specification, applicable Test Reports (One through Four), subsequent Engineering Change Notices (ECNs), and Modification Design Change Notices (MDCNs).
The inspectors found that appropriate design inputs from codes and standards, and applicable design criteria were identified and appropriately considered.
Required technical, design verification, and independent design reviews were performed.
Post-modification test procedures focussed on the installed changes and the appropriate acceptance criteria were established to verify proper functional requirements, including voltage values for the modification.
2.1.2 Service Water Fouling - Phase 3 - EWR No. 4658C This modification involved the third phase of the service water fouling modification project.
Part 1 included the provision for larger drain lines/valves to afford an improved flushing capability for the auxiliary feedwater pump suction piping.
Parts 2 and 3 had an initial operational test to verify correct installation and operation prior to a final preoperational system functional test.
The inspector concluded that the modification was well engineered and included good design inputs.
2.1.3 Amptector Upgrade - EWR No. 4225 EWR No. 4225 is a modification planned for installation during the upcoming 1995 outage.
This modification replaces the electro-mechanical trip mechanisms in DB-type breakers with a solid state amptector device.
This new device willupgrade the overcurrent protection and allow for response testin Review by the inspectors of the completed design analysis found that assumptions used were conservative and technically reasonable.
The inspectors determined that proper evaluations were made for the specific application of amptector devices for achieving optimal protection of plant loads and coordination with feeder breakers.
2.2 Main Feedwater Regulating Valve (MFRV) Trim Change The previous trim design within the MFRVs resulted in the valves being about 40-50% open at full power.
This was significantly less than the original valve design desired operating range of approximately 85% open, and resulted in the MFRV being operated in the steepest part of its Cv versus
% open curve.
This position resulted in flow oscillations with the feedwater control system in automatic, and possibly contributed to the plant trips experienced in November 1993 and April 1994.
I, Qi During the past refueling outage, the licensee installed a new MFRV trim design that will permit the MFRVs to be positioned close to the original full power operation design point of 85% open.
Engineering personnel at the plant prepared Technical Staff Request/Technical Evaluation (TSR/TE)93-561 to describe and install this new trim in.the MFRVs. The inspector reviewed TSR/TE 93-561, including the various technical documents referenced therein, and had the following comments:
Mechanical Engineering performed Design Analysis No. DA-ME-93-123, "Evaluation of MFRV Differential Pressure at Normal Plant Operation."
This engineering work was needed to define the technical requirements for the procurement specification.
The responsible engineer performed a comprehensive design analysis to develop a new MFRV control data sheet for use by the valve vendor (Copes-Vulcan) in designing the new valve trim.
2.
In TSR/TE 93-561, the licensee documented a weight discrepancy regarding the overall MFRV assembly that had been identified by Mechanical Engineering.
The actual valve/actuator assembly weight was determined to be significantly greater (3885 lbs. to 4625 lbs. for a 19% increase)
than the weight used in the last detailed stress analysis for the applicable piping installations.
This weight discrepancy was primarily attributed to a MFRV actuator change from the original Copes-Vulcan version to a Fisher Model 473-1-4-5, which added 500 lbs.
The licensee considered the 19%
weight increase acceptable for the purposes of TE 93-561 since it was less than a 20%
guideline included in EPRI Report NP-5639, "Guidelines for Piping System Reconciliations," Revision 1, May 1988.
In light of the fact that the MFRVs are installed in high energy lines, the inspector questioned the licensee's use of a 1%
margin in the absence of an updated piping stress analysis.
I The inspector further noted that modifications controlled via the EWR process, such as EWR No. 4658C discussed in Section 2.1.3., are more tightly controlled and reanalysis work is normally completed before a modification is installed.
Mechanical
Engineering noted that they had initiated Identified Deficiency Report (IDR) 015-94
'n February 24, 1994, to perform an updated piping system stress analysis.
This engineering work was in progress during this inspection and was targeted for completion by June 1, 1994.
The licensee indicated that the analysis for the first system was basically completed and that the piping stresses using the updated loads were well within code allowable values.
The licensee expects the overall analysis to be completed by June 30, 1994.
3.
Reactor plant trips due to MFRV problems occurred in November 1993 (described in Licensee Event Report 93-006) and in April 1994.
Both events were attributed to problems with the Bailey Model AV1 positioner, which is the control device installed to monitor MFRV stem movement, and provide an air pressure signal to the actuator for controlling MFRV movement.
The April 1994 event involved a stuck pilot valve assembly in the positioner for the "A" MFRV. The licensee issued Nonconformance Report (NCR)94-084 for corrective action.
The licensee corrected the immediate problem by installing Bailey Model 5321030 positioners for both MFRVs that had been used previously until they were replaced in 1991.
The licensee closed out NCR 94-084 based on this corrective action.
However, this NCR closure may have been premature since the long term resolution of the AV1 positioner failures had not been completed.
The licensee noted that. this action would be included as a part of Root Cause Analysis M94-007, which was still open to address the April 1994 "A" FWRV failure.
It was apparent to the inspector that there was a possible lack of licensee management oversight in the implementation of TSR/TE 93-561.
None of the documentation reviewed by the inspector associated with this TSR/TE involved the plant onsite review committee (PORC). Ifthis is the case, this would be inconsistent with PORC's responsibilities, especially in light of the recent plant trips due to MFRV problems.
The inspector considered that this may be an example of improper use of the TSR/TE process to make changes to the plant.
Pending further actions to resolve Items 2, 3, and 4 above, this item is unresolved.
This includes satisfactory completion of the MFRV piping stress analysis, long term corrective actions for the MFRV AV1 positioner failures, and additional review of other examples of the use of the TSR/TE process for assurance of adequate licensee management oversight (Unresolved Item No. 50-244/94-15-01).
2.3 Throttling Component Cooling Water (CCW) to Residual Heat Removal Heat Exchangers The licensee determined that the potential for flow-induced vibration existed for flows in excess of 2500 gpm through the CCW heat exchangers (HXs). This information was based on studies and correspondence from the HX vendor (Atlas Industrial).
Design flow given in
the UFSAR is 2780 gpm.
The licensee identified this concern during the past refueling outage by adjusting the flow control valves (CCW 780 A/B) to limitthe flow to 2500 gpm to the CCW HXs.
The inspectors reviewed Safety Evaluation 1011, Revision 1, and the various references noted therein that evaluated the impact of these changes on plant operation while at power.
The licensee performed a thorough design analysis (DA-ME-93-0052, "CCW HX Flow Analysis for Potential Flow Induced Vibration") to evaluate the potential for flow-induced vibration.
The actual valve position for CCW 780 A/B was changed from 35 degrees open to 30 degrees open.
The inspector verified that plant procedures were appropriately changed, to reflect the new valve positions.
The licensee further indicated that FSAR changes were being processed to reflect the reduced flows and the resultant lower cooldown rates.
The licensee had inspected both CCW HXs and plugged less than two percent of the total number of tubes.
The inspectors also noted that the licensee had submitted a 30-day written report, in accordance with 10 CFR 21, informing the NRC of this potential component defect.
The inspectors concluded that the licensee performed good evaluations and took appropriate actions concerning the potential flow-induced vibration problems for.the CCW HXs.
2.4 Engineering Work Backlog The inspector discussed with engineering management the licensee's efforts regarding the definition and control of the engineering work backlog.
The licensee has issued a new engineering procedure to prioritize all engineering work items.
The licensee's intent is for all personnel to use this procedure so that engineering work items are systematically defined and scheduled on an ongoing basis.
The licensee was not as far along in developing an engineering work tracking system that would enable them to correlate engineering work performance to engineering work expectations included in the engineering department business plan.
Draft portions of this engineering work tracking system were available, but not sufficiently developed for the inspector to review.
2.5 Configuration Management Information System (CMIS)
The inspectors reviewed the.licensee's methodology and development of an information infrastructure designed to support configuration management.
This review was made to understand how baseline configurations and design basis information is established and planned to be maintained.
In 1988, RG&E began a program improvement project for the storage and retrieval of configuration management information.
The licensee initiated this project in recognition of a growing industry-wide awareness that this information is essential for ensuring that the plant willfunction as originally designed and subsequently modified.
The purpose of this program improvement project was to ensure processes are developed and implemented to maintain configuration and design basis information, ensure the plant is modified and operated within
the design bases, and assure plant documentation accurately represents the plant hardware.
The inspectors reviewed this area to understand the process for assembling this information and status of this project.
The licensee began this project with over one hundred paper or electronic databases and no central controlled system.
The effort to group collectively these databases resulted in the integration into three major databases.
This collection was to provide continuous improvement for controlling information during stages of development.
At the time of this inspection, three databases existed and initial training had been provided to personnel for use of information, familiarity, and understanding the functionality of the database.
Additional training is planned following final system implementation of a single integrated information system.
Maintenance of the completely integrated system is scheduled to begin in August 1995.
This system is planned to provide cross-referencing with design basis documents.
The system is designed to provide sufficiently detailed design information to support accomplishment of modifications, with a clear understanding of the affect of the modification on the system functions or system interfaces.
Systematic reviews were performed of the configuration walkdown information and compared to design basis information.
This comparison was to ensure the baseline plant configuration was consistent with plant design bases.
Necessary actions have been developed to provide users with reliable data and training regarding the importance of following formal practices for using and maintaining accuracy of configuration management information.
The inspectors concluded that RG&E's effort to develop and maintain processes for integrating many information management systems in a single controlled source was good.
The controls and verifications developed to establish and maintain plant configuration were found to be effective for ensuring utilization of proper and current configuration information.
2.6 Design Basis Documentation The inspectors reviewed RG&E's means for maintaining information that identifies the specific functions to be performed by a structure, system, or component, and the specific values or range of values chosen for controlling parameters as reference bounds for design.
This review was performed to verify that information is maintained in accordance with the original plant design.
Design basis information for structures, systems, and components was contained within many documents.
These documents include the operating license and its associated conditions, the updated final safety analysis report (UFSAR), technical specifications (TS) and bases, written NRC safety evaluations and correspondence referenced therein, RG&E correspondence submitted to the NRC in support of the approved license, and TS amendments or source
documents in the UFSAR.
Plant modifications under consideration require an evaluation by Engineering of the effect on existing design bases for the proposed change to ensure that appropriate design margins and safety system functionality are maintained, and that unnecessary challenges to safety systems willbe avoided.
Ginna Station has three design basis documents (DBDs) completed.
These DBDs are single-source controlled information documents that collate the design basis for a particular system.
DBDs present functional descriptions, interfaces, and equipment for selected systems and topical areas.
The completed Ginna DBDs are for the safety injection, auxiliary feedwater, and instrument air systems.
In addition, three Westinghouse DBDs that envelope generic two loop plants including Ginna, Point Beach, and Prairie Island, have been procured and were being updated to reflect Ginna specific features at the time of this inspection.
In addition to the existing DBDs, RG&E has initiated a DBD Project to develop the-capability to retrieve the licensing basis and design basis supporting source documents associated with all systems and selected topical areas.
This initiative was developed to deliver a controlled information source defining these bases and supporting information licensee's Business Plan projects a project completion date by December 1996.
RG&E
'management has committed substantial resources to support this project.
This initiative was being performed under EWR No. 4895D.
Further actions being taken by RG&E include the development of system descriptions and engineering procedures.
These system descriptions are planned to utilize information described in QUAD-7-89-007, "System Primary Function Analysis Report for Ginna Nuclear Power Station Equipment Safety Classification," and QUAD-1-89-012, "System Functional Report," to determine any potential impact as a result of a proposed change to a component or system utilized to mitigate accidents or transients.
These system descriptions are planned to evolve into the DBDs.
The inspectors concluded that this project exhibited a good engineering initiative. The creation of DBDs establishes controls for performing plant modifications, safety reviews and evaluations, and operability and reportability evaluations.
2.7 Self-Assessments of Engineering Activities The inspectors reviewed the adequacy of the licensee's most recent self-assessment pertaining to the modification control process.
This review was performed to determine the appropriateness of the scope of the assessment, the quality and disposition of findings, and the follow-up of any corrective actions taken based on deficiencies identified in the audit report.
Audit Report No. 93-37: TGT examined a selection of modifications performed during the 1992 outage and a selection planned for the 1993 outage.
The review by the licensee
O.
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evaluated all facets of the modification process, with particular emphasis on the performance of 10 CFR 50.59 safety reviews and EWR packages and procedures.
This review did not identify any programmatic problems or significant findings.
Observations presented were clear and understandable.
The inspectors concluded that this self-assessment performed by RG&E was adequate in scope to address pertinent aspects of the modification process.
3.0 CONCLUSION The inspector concluded that the design change and modification program was being controlled well with clearly detailed administrative procedures.
Based on the modification packages reviewed, the engineering products provided the necessary guidance to safely modify the plant.
Information was not yet available regarding several concerns related to MFRV trim changes that resulted in an unresolved item.
4.0 MANAGEMENTMEETING The scope and purpose of the inspection were discussed at an entrance meeting conducted on June 6, 1994.
During the course of the inspection, the inspectors'indings were discussed with the licensee representatives listed in Attachment 1.
An exit was conducted on June 10, 1994, at which time the preliminary findings were presented.
The licensee acknowledged the findings and conclusions, with no exceptions taken.
Further, the bases for the preliminary conclusions did not involve proprietary information, nor was any such information discussed or expected to be included as part of the written inspection report.
Attachment:
Persons Contacted
'
ATTACHMENT1 Persons Contacted R
he ter s an Elec ri om an
- C. Anderson R. Arnold P. Brown
- J. Dunne
- J. Gashlin F. Gilbert
- T. Harding
- G. Hermes
- M. Kennedy
- S. Lawlor
- D. Markowski
- ¹ Oliva
- P. Swift
- C. Vitali E. Voci
- T. Werner P. Wilkens
- G. Wrobel QA Manager Lead Engineer, Design Basis Project QA Engineer, Technical Processes Mechanical Engineer Construction Engineer Mechanical Engineer Lead Engineer, Modifications/Special Projects Lead Engineer, Safety and Licensing Director, Configuration Management Project Associate Mechanical Engineer Lead Mechanical Engineer Senior Electrical Engineer Electrical Engineer Lead Mechanical Engineer Manager, Mechanical Engineering Director, Technical Processes Department Manager, Nuclear Engineering Services Manager, Nuclear Safety and Licensing
- Indicates those in attendance at the exit meeting held on June 10, 199 li
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