IR 05000244/1992001
| ML17262A805 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/05/1992 |
| From: | Gray E, Gregg H, Lohmeier A, Roy Mathew NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17262A801 | List: |
| References | |
| 50-244-92-01, 50-244-92-1, NUDOCS 9204070196 | |
| Download: ML17262A805 (40) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
2 2 R
.
22LRRPf Docket No.
50-244 License No.
DRP-18 Licensee:
Rochester a
and Electric o
r ti n dh fp
. Rochester New York 1464 Facility Name:
Ginna Nuclear Powe'r Plant GNPP Inspection at:
NES Office and GNPP Site p
C d d'~22 Inspector:
R. Mat ew, Reactor En ineer Electrical Section, EB, DRS date H. Gregg, S
.
eactor Engineer, Systems Section, EB, DRS date A. Lohmeier, Sr. Reactor Engineer, Materials Section, EB, DRS dat Approved by:
E. H. Gray, Chief, Materials Section, Engineering Branch, DRS 3.5 date hhhd:
Th p
fhi
p
<<2 d
2 <<h ff of the engineering and technical support activity at the Ginna nuclear power plant onsite and offsite offices.
Considered in the inspection were the engineering organizations, corporate engineering management, engineering process developments, and design and facility modifications.
'2t204070196 920330 PDR ADOCK 05000244 G
Clb
Re@i~it:
Review of the inspection findings indicated Ginna nuclear power plant to have an engineering organization with a goal oriented management, a well planned engineering
. process development program, and the ability to implement design and facility modifications in a manner commensurate with that expected in a safe and reliable nuclear plant.
A violation and two associated unresolved items were found by the inspectors related to environmental qualification of motor operated valves.
These issues occurred in 1990 and were not indicative of present engineering capabilit Ck
1.0 S
of In tion The scope of this inspection was to review and evaluate the engineering and technical
.support activity at Ginna Nuclear Power Plant of Rochester Gas and Electric (RG&E)
Corporation to ascertain that appropriate engineering effort is directed toward operation of the plant within its design basis and in compliance with Technical Specifications, Final Safety Analysis Report, and Code of Federal Regulations.
Included in this inspection was examination of those activities directed at improvement of engineering performance.'his included consideration of staff levels and organization, communication, engineering performance evaluation, modification program implementation, problem anticipation, identification and appraisal, facility design change implementation, and management support of initiatives to improve plant performance and safety.
2.0 Findin s - En ineerin and Technical u
rt
2.1 o
rate Mana emen Ginna Nuclear Power Plant supplies half the Rochester Gas and Electric Corporation (RG&E) customers'lectricity requirements, The financial performance of RG&E is highly dependent on. the sustained periods of Ginna safe and reliable operation.
An availability factor of over 82% achieved by Ginna compares favorably with other nuclear power plants throughout the nation.
In the business plan for 1992-1995, Nuclear Production Division management has provided a mission statement from which long term goals for the division are derived.
These include quality, public safety, personnel safety, regulation compliance, employee achievement, reliability, and economy.
Key performance indicators are specific expected performance goals over each year.
from which the performance of each functional area can be measured.
Review of this management system by the inspector indicates that a comprehensive, mission oriented system of goals and performance indicators has been established at Ginna for the Nuclear Production Division which provides for effective management of the engineering, technical support, and other functional activitie.2 En ineerin r
ni i n The Vice President of Ginna Nuclear Production Division reports to the President and Chief Operating Office through the Senior Vice President of Production and Engineering.
Engineering functions exist within both the site and Nuclear Engineering Services Department.
There is. also a separate engineering department for the steam generator replacement project, since this is a long range major facility improvement activity. The engineering function is divided between site engineering, with direct activities reporting to the plant manager, and engineering reporting to the offsite nuclear engineering services department manager.
The format of the engineering organization at both the site and offsite is functional, with the exception of special cases of major projects, such as the steam generator replacement.
The inspector noted that the engineering'organization functions well, with good communication between engineers at the site'and those of the offsite services organizations.
Ginna is an earlier nuclear facility and this type of organization is appropriate to its size.
2.3 Nuclear ervi En ineerin Engineering personnel were found to be knowledgeable and technically competent in the implementation of modifications and in each of the engineering issues reviewed during the inspection.
The cognizant engineer for each modification was at the site on a full time basis during the outage and the licensee's efforts to improve planning and to release work packages early has been reflected in the effective results achieved.
Present staffing of 111 authorized full time RGE engineering positions with 100 now in place represents an increase of some 12 engineers in the past year and a half.
The increased staff has provided needed work resources that contributed to productivity. The new engineers, with experience from other, nuclear plants, have provided a renewed emphasis to various programs, such-as probability risk assessment (PRA), master equipment and Q-listing compilation, computer database indices, engineering modifications, and'the support of site activities.
It was also noted that there is little turnover of engineering personnel and use of contract personnel is minimal.
However, qualified contract engineering resources are utilized when needed.
Increased staffing, with emphasis on obtaining cross functional experience, has improved the engineering organizatio Several work activities were reviewed to determine the effectiveness of engineering participation in plant activities.
The review determined that the corporate engineering group actively participates-in the daily meetings at the site which are held to discuss the plant status and to expedite the needed engineering support.
The active participation of management representatives from different organizations in these meetings provided for an effective interface between engineering and plant organizations.
The site technical engineering staff and corporate engineering staff actively participate in the disposition of non-conformance reports (NCRs) and root cause investigation.
The personnel to complete these tasks are assigned by their respective functional department managers.
Licensee engineering developed a 10CFR 50.59 safety evaluation/screening guideline for performing activities other than minor and major modifications, These guidelines were incorporated in the corporate and plant procedures.
The 10CFR 50.59 review for modifications and NCRs was found to be thorough.
'he inspector noted that the licensee has a Potential Conditions Adverse to Quality (PCAQ) program which identifies and resolves safety significant issues with appropriate engineering and plant personnel review.
The review of a sample PCAQ indicated that potential safety concerns were properly identified and directed for resolution.
Operability review is performed by plant personnel in accordance with plant procedure A25.1 for all potential safety concerns.,No unacceptable conditions were identified during this review.
Root cause determination for LERs and potential safety issues are coordinated and documented by the corrective action report (CARs) group.
The sample review of open CARs indicated that CARs were closed out in a timely manner and root cause investigation is completed with engineering participation and evaluation.
2.4 ualit Performance Indicator I
RG&E Nuclear Services department has an extensive system of quality performance indicators which are tracked on a monthly basis.
Included in these indicators are corrective action reports, engineering change notices, erosion/corrosion wall thickness reports, field change requests, identifi'ed deficiency reports, nonconformance reports, engineering nonconformance corrective action, nuclear assurance surveillance reports, potential conditions adverse to quality, technical support requests, 10CFR 21 evaluations, 10CFR 50.49 evaluations, procedures, and trainin.5 En ineerin Through this computerized reporting system, engineering management has an effective means of oversight of the engineering organization performance.
lit A s nce Au it Quality performance at Ginna is directed by a department manager reporting to the senior vice president of production and engineering.
Reporting to the quality performance department manager are the quality assurance manager, the nuclear assurance manager and the senior performance analyst.
The quality assurance manager is responsible for audit, procurement, design and surveillance activities.
The nuclear assurance manager is responsible for the quality control activities at the site.
The inspectors reviewed selected QA audits (Nos. 91-35 and 90-47) performed on engineering activities.
The purpose was to ascertain the degree to which the audits addressed the functions of the engineering groups, the significance of the findings and the licensee's corrective actions and response to the.
findings.
The audits identified minor weaknesses in the administrative control of documents and several weaknesses in controlling quality assurance
.
documents.
A review of the bi-monthly status report for the period ending October 31, 1991, indicated that the open action items to be resolved by the engineering management were minimal and the licensee was responsive in resolving the audit findings.
The inspectors concluded that quality assurance involvement in monitoring engineering effectiveness was adequate.
2.6 mmni in Effective communication exists between plant and engineering personnel at RG&E, This was evidenced by the presence of corporate engineering staff at the site to support the engineering and technical support needs of the plant.
The licensee has established morning and afternoon meetings to discuss plant activities pertaining to design, operation, and maintenance action items.
Furthermore, the outage and major modification planning committee, along with corporate and site technical engineering support, plan and schedule the tasks for the upcoming outage through regular meetings conducted on a weekly and monthly basis.
The review of a major modification (digital feedwater control system) indicated that good communication existed between various engineering and support organizations for the successful design and operation of the syste.7 En in rin P
'
i n in Emer enc Pl nnin The inspector reviewed the role played by engineering during plant emergencies.
Engineering personnel have been trained in emergency procedures and play an active role in the support of the site during emergency drills. Specific roles have been assigned to engineering personnel for emergency events.
Senior management (or designated alternates) personnel carry pagers to communicate with them.
There are two event rehearsal drills each year in which engineering plays an active role.
Computer technology provides for instant communication of event information.
3.0 n ineerin Pr s Devel ments 3.3 de 3.1.1 Pi in and Instrument Drawin s
&ID WR 33 7 A followup was made of the program that was completed with the release of 147 riew P&IDs late in the last SALP period in order to determine how the licensee dealt with post release problems that normally occur.
The inspector determined that engineering maintains effective control of the P&ID drawing program by:
1) updating drawing punch list items and improving the drawing interface arrow notations, 2) preparing drawings of 19 skid mounted piping arrangements and 19 fire sprinkler drawings and the assigning of plant valve numbers to the skid mounted drawings, 3) contracting with Gilbert Commonwealth, the original design agent, to review the drawings, and 4) incorporation of the design change requests (DCRs) that were in process during the program release.
The inspector also determined that training on the use of the new P&IDs was given to the corporate Nuclear Engineering Services and site technical staf.1.2 ru tu l Arran em n D win rade The in-house drawing program to reformat and improve the plant structural arrangement drawings was initiated and completed in 1991.
The drawing upgrade program improved drawing legibility and provided for dimensions of the plant as it exists today from walkdown inspections and document searches of engineering work requests.
Forty (40) structural arrangement drawings were completely redrawn in a new format that provides clear identifier balloons for the equipment shown.
A master index list drawing and 45 new drawings that reference equipment. at various building eleyations were also developed to be'used in conjunction with the arrangement drawings, The inspector, examined the new drawings and found them to be improved over the old drawings, many of which were illegible or not up to date.
The inspector further verified the licensee's intent and method to be used in maintaining the new drawings current.
The inspector determined that the new drawings are classified now under the configuration management program as type "B" controlled drawmgs.
The licensee's drawing upgrade efforts have been effective as evidenced by the P&ID and structural drawing upgrades.
The use of computer aided drawing (CAD) techniques have enhanced the drawing upgrade projects.
3.2 Desi n Ba i Doc m nt BD Rec n tituti n The Nuclear Engineering Services Department is in a comprehensive foundation building stage of the DBD reconstitution program.
Good progress has been made in beginning this work to'enable an effective'end product to be complete by the end'of 1994.
Three experimental pilot DBDs relating to safety injection, instrument air, and auxiliary feedwater were done in 1990.
They were performed by three different contractors and resulted in three varied approaches that did not satisfy the licensee's needs.
These DBDs were not formally accepted nor were they fully placed in use. However, they were useful in the planning of how to achieve a good DBD. Lessons learned through these three pilots were:
1) the end product should be placed on an electronic data base, 2) there is a need to gather all information needed to substantiate the equipment and system data to facilitate the preparation of DBDs, and 3) the use and maintenance of the DBD end products must be addressed prior to their developmen The licensee has been addressing the issues learned from the pilot DBD program and by several concurrent activities.
A massive electronic indexing and assembling of all NRC/Ginna correspondence including attachments was devised by engineering and completed in late 1991.
Recall of information can be obtained by system, topic, or equipment type menus.
This data base has been a valuable engineering tool. In conjunction with the two loop Westinghouse owners group (2LWOG), three DBDs are being developed.
These include the reactor coolant system, chemical and volume control system, and reactor protection systems.
These DBDs have the benefit of the pilot experiences, with more precise form and content requirements, and the combined thinking of three licensees.
The first of the 2LWOG DBDs is complete in draft form for review and the other two willbe completed in the first half of 1992.
Concurrently, the Westinghouse owners group (WOG) has agreed to provide for the electronic indexing of the residual heat removal system.
Consideration willbe given to index additional systems in February 1992.
The licensee's engineering staff has also developed the electronic indexing of two pilot engineering modification work requests that are to be followed with a massive indexing effort for all engineering work requests (EWRs).
In addition, electronic indexing of the architect/engineer documentation is to be performed.
'he present DBD plan includes completion of indexing, obtaining user input, developing the final writers guide, in-house reworking and acceptance of the 2LWOG and the pilot DBDs, review of the overall matrix of systems and topical DBDs to be performed, and completion of DBDs by the end of 1994.
This plan is currently being reviewed by licensee management.
/
The inspector concluded that the licensee's effort in DBDs has been comprehensive and the development of electronic indexing of information will provide the foundation to enable effective DBD output.
The electronic information data base approach willprovide additional benefits in preparing DBDs.
Although there are not finalized DBDs in place, the engineering methodology has been a substantial effort that can result in an improved quality end product.
3.3 Pr ili tic Risk A es ment RA The PRA project committed to by the licensee to provide a level 1 and level 2 PRA is an aggressive program that goes beyond generic letter requirements.
Excellent progress has been made and recent in-house staff additions of experienced PRA personnel have been evident.
The PRA is being developed under QA program controls and placed in a computer data base to afford ready access and change capabilit Review by the inspector of the data analysis work package related to component reliability parameters indicated that a comprehensive program was in progress that considers the number of component failures vs. number of component demands and'he time which the component was in operation or standby.
Twenty (20) plant systems were included in this effort. This program provides for data analysis to be utilized in probabilistic risk assessment directed at Ginna component reliability. This project is believed to be an indication of the ability of Ginna engineering to conduct the state-of-the art evaluations necessary in assessment of nuclear power plant safety and reliability.
3.4 Drawin han e R ue CR The new drawing change request process developed by engineering and implemented in December 1990 has been effective.
Complete revisions to procedures QE-324, "Preparation Review and Disposition of Drawing Change Requests" and QE-303, "Preparation Review and Approval of Engineering Drawings," provided clear instructions, simplified the change process approval, redefined turn around times to complete changes, clarified different drawing types, identified persons responsible, and indicated how the process would be controlled.
In conjunction with the procedure upgrade, a new DCR tracking log status was instituted.
The tracking log is issued monthly and provides details of the change, responsible personnel, dates assigned, posting locations, and close out dates.
The new DCR process and tracking system, significantly improve the capability to make and control changes.
, 3.5 En ineerin W rk R e t WR A review was made of the EWR status.
It was noted that the current backlog as of 1/9/92 was slightly reduced to 449 open items as compared with 484 items a year ago.
Although 99 items were closed through completion or elimination, the change over the past year did not show a significant decrease.
The current listing of open EWRs is routinely reviewed by engineering to ensure that safety issues are completed.
The backlog is basically a paperwork issue that must be addressed.
The backlog issue was discussed with cognizant engineering personnel and engineering management and there is an awareness of the problem.
The engineering process upgrade program focus items, 1) work identification and 2) managing work priorities, are scheduled to be complete by first half of 1992 and may aid in resolving this problem.
At present, there is no quantitative goal for specific backlog reductio The problem of EWR backlog reduction was discussed at the exit meeting and the licensee's'anagement committed to address this problem.
3.6 P'he licensee has established a viable system for controlling the engineering work load and for establishing and revising priorities.
Procedure guideline No. OMG-2 delineates the purpose and guidelines for the integrated prioritization of modifications and activities to recognize the needs of RG&E as operator of the plant.
This is accomplished through a continuing process of selecting, integrating, prioritizing, and scheduling plant improvement activities on the basis of safety, regulatory'onsiderations, plant improvement,= plant operability, satisfaction of corporate goals, and optimization of the allocation of resources.
The six categories mentioned above have been selected as the attributes for defining priority of projects.
These six attributes are further divided into sub-topics which would aid and simplify the evaluation rating process.
The priorities are determined by a scoring system with numbers 0-10 and attribute multipliers to determine the final score.
Allattributes do not have equal importance.
Safety has the highest multiplier and corporate goals the least.
The priority for engineering work load is approved by the planning committee with department managers from various engineering and other support group organizations.
Modifications are assigned, prioritized and tracked, periodically.
The system for-assigning priorities to plant modifications
'as been given the proper safety perspective.
Projects are planned for the future five year period and the items are listed as pre-outage or outage activities.
The priority system is influenced by the budgetary process and management makes the final determination of allocating funds for the modifications.
To improve the effectiveness and efficiency of the engineering services, a
process upgrade program was initiated in June 1991.
One of the focus areas in that program is to analyze the existing prioritization program and develop a rational process for identifying Nuclear Engineering Services priorities.
3.7 n ln rin Ini'v 3.7.1 En in rin Proces radin Pro ram In an attempt to improve the effectiveness and efficiency of engineering and technical services, the NES department initiated a process upgrade program.
The program provides for clear statements of objectives and scope for ten (10)
focus areas including work identification, management of work priorities, design planning, procedure upgrading, project planning and management, implementation, document coritrol, resource utilization, NES support, and
performance management.
This program has an organizational structure and target schedule and is a comprehensive effort to direct activities toward improved engineering performance at Ginna.
3.7.2 nfi
'
M a
ment Pr ram The,.RG&E configuration management program is a longer range project designed to improve engineering performance through several specific categories including Q-list project, maintenance procedure upgrading, setpoint verification program, P&ID program, vendor technical manual program, design basis document development, and calibration procedures, The four (4) key elements of this program are baseline configuration development, design basis development, change control procedures and information access.
These programs are consistent with the general improvement of the engineering process throughout the nuclear power generation industry and Ginna engineering is effectively managing this improvement effort.
4.0 De i n M ifi tion P cka e Review 4.1 EWR 47
- IST heck Valve Testin Additions ASME Section XI valve leak testing and assurance that check valves fully open to pass design flow required the installation of test connections and flow instrumentation.
This eliminates the need for full open verification through disassembly of the valve and the associated problems of radiation exposure and proper reassembly.
Prior NRC findings and the licensee's self review of IST check valve testing methodology provided further motivation for this modification.
The inspector reviewed the specific design criteria, design verification, and safety analysis details of the flow instrument additions downstream of check valve 8655 in the spent fuel pool pump "A" discharge piping and downstream of check valve 4023 in the turbine driven auxiliary feedwater pump recirculation piping.
Each aspect of the modification was fully described, including performance requirements, code and regulatory issues, and
.
installation and testing details.
The inspector also verified the design analysis calculations for the tubing and supports to be installed and found the stress levels to be acceptable.
The design analysis calculation methodology was well organized and displayed a good engineering produc The inspector reviewed the first test results of the auxiliary feedwater recirculation flow and concluded, the modification provided a simplified and more accurate system of flow determination and-verification of check valve full open position.
Prior concerns, therefore, have been addressed.
-.4.2 EWR.47-1 Adv n Di i I F w ter M hni I
ni in The advanced digital feedwater control system was a major modification effort that was completed during the 1991 outage.
Extensive tubing and tubing support additions and upgrading to current seismic standards required a substantial mechanical engineering output.
To accomplish the objectives of this modification, engineering developed in-house a standard tubing support design analysis (ME 90-009) for sixteen predesigned standard supports.
In the design of each, the predesigned supports analyses and calculations for load capacity was developed.
A series of well prepared drawings (DWG 33013-2458 and 2459) depicting each of the standard supports was produced for fabrication and assembly of the supports, including their location.
In conjunction with the predesigned support effort, an extensive analysis and release of generic support spacing information was also developed in-house (EWR 4218-TM1).
This design analysis compilation provides the technical basis for determining generic span lengths, offsets and loads, and includes varied tubing sizes and materials.
The-resulting output of this well planned effort is a series of
, tab'ular listings of maximum span lengths for various location elevations and conditions including both dead weight and fluid filled weight on two types of simply supported beam configurations.
The inspector reviewed the calculations for support configuration for mounting of two gauges (TP41) and for mounting of a two'tube clamp to a horizontal structural steel member.
No problems were identified by the inspector.
The engineering associated with this modification was forward'thinking and the predesigned supports and generic spacing was an engineering initiative that has contributed to the successful modification completion.
Also, the predesigned supports and generic spacing designs are being utilized in other modification.2.2 Electrical onsiderati ns Modification package EWR 4773 provided documentation for replacing the Foxboro Analogy Feedwater Control System with the Advanced Digital Feedwater Control System (ADFCS) to control reactor power from 3% up to full power.
The new system is a Westinghouse digital/microprocessor, feedwater control system.
The design will provide stable, automatic, continuous, sequential split range operation and transfer of control between the main and bypass feedwater control valves.
This modification was initiated by the licensee to eliminate or greatly reduce reactor trips associated with single component failures and steam'enerator low-low levels at low power.
The inspector reviewed the historical background, safety features, design and design change process for this modification.
The review indicated the following: Westinghouse owner's group trip reduction and assessment program concluded that 38% of all Westinghouse PWR reactor trips were due to feedwater- (FW) control problems.
Ginna had two plant trips (May and June 1990) caused by FW control malfunctions.
The licensee assigned highest priority for completing this modification.
The engineering effort began in August 1989 and engineering design verifications and integrated assessments were completed in February 1991.
The licensee staff visited the supplier facility (Westinghouse)
and verified the ADFCS installation and operation of the unit installed at Prairie Island Nuclear Station, Wisconsin.
During the conceptual design review, the design group effectively interacted with the users of this modification and incorporated all their comments and input. A modification follow group consisting of several personnel from different organizations were responsible for completing this modification. A formal meeting was scheduled regularly to monitor and to provide recommendations for the implementation of the modification. The inspectors interviewed several personnel to determine their understanding of the modification.
The review determined that effective communication existed among the various engineering and technical support organizations and the personnel were knowledgeable.
The operators interviewed revealed that this modification significantly improved the operating/plant startup proces The design verification, safety evaluation, design input & design analysis reviewed were thorough and in accordance with plant procedures.
The inspectors noted that a Technical Specification change
.request was processed by the licensee to revise the requirements of the Technical Specifications to eliminate the steam flow/feed flow mismatch reactor trip. The purpose of the steam flow/feed flow mismatch reactor trip was to meet IEEE 279 requirement to prevent
interaction between control and protection functions.
This amendment
'as approved by the NRC.
The inspector noted that the median signal selector (MSS) was used to justify elimination of the low feedwater flow reactor trip function in addition to enhancing fault tolerance to input signals to meet the IEEE 279 requirements.
It was also observed that process inputs to ADFCS increased from 7 inputs to 40 inputs and electrical separation/isolation and redundancy were maintained for the electrical/instrumentation components and circuits.
The licensee installed other modifications and changes during the implementation of this modification.
For example, 1) four channels of steam generator wide range level instrumentation were installed with qualified instruments for post accident monitoring, 2) identified voltage regulation concern in the instrument bus distribution system regarding input voltage requirements for Foxboro power supplies.
NCR No.91-415 and potential conditions adverse to quality report No.91-027 were issued to address this issue, and 3) relocation of main control board instruments to enhance the human factor aspects of the modification.
The licensee stated they are planning two future modifications to eliminate unnecessary alarms due to noise spikes and replace turbine trip interlocks.
The inspectors performed a walkdown to verify the electrical and instrument action portion of the modification.
No unacceptable conditions were identified during this review.
The inspectors concluded that the modification was well planned, good interface/communication existed, and good engineering design and project experience were utilized for the successful completion of this modificatio.3 EWR4 T
n ferD N n-1E ad The inspector's review indicated that the station vital batteries are appropriately sized.
However, the licensee has an initiative to increase the
.
existing design and capacity margin of the batteries.
As a result, non-1E dc loads from the vital batteries were transferred to the TSC batteries.
The transferred loads included the turbine dc lube oil pump motor, the air side seal oil backup pump, and the two circulating water discharge motor operated valves.
The design input, design verification, and safety analyses contained appropriate requirements.
The affected battery sizing and loading calculations were revised to-reflect the as-built conditions.
No unacceptable conditions were identified during this review.
4.4 Diesel enerator Buildin M ification EWR
The inspector reviewed the Ginna Station Structural Program in response to the Systematic Evaluation Program undertaken by NRC in 1977.
The program included "hardening" of the Diesel Generator Building (DGB) to meet the requirements resulting from extensive seismic analyses performed on the main structural systems at Ginna Station.
Consideration in the design upgrading included the effects of wind (tornado), snow, earthquake (OBE, SSE), and flooding.
Included in the inspector's review were design requirements, safety analysis, design review, and analyses'odels.
A walkdown of the completed building modification by the inspector provided for visual verification of this building modification project.
Review of this modification program by the inspector indicated the capability of the licensee to effectively manage their activity and those of retained consultants in satisfaction of new safety requirements at Ginna in response to NRC safety evaluation results.
The modification of the building was carried on by Ginna.
4.5 team enerat r Re lacement Pro ram The inspector reviewed the Ginna program to consider replacement of the steam generators.
The program consisted of one piece vs. two piece replacement study, review of replacement techniques, auxiliary support facilities, steam generator bid evaluation, optical templating (positioning study), and replacement approva The bid review is intended to be completed in March 1992, and the steam generators ordered in May 1992, with replacement targeted for 1996 or beyond.
The activity of Ginna in effective management of this project was found to be commensurate with demands of a safe and reliable nuclear power generation system.
This activity included establishment of-a separate project group and participation with other utilities in preparing a specification based on the overall of industry steam generator operating experience.
The decision to replace the steam generators willbe based on the best available technology together with economic evaluations.
4.6 Environmental u lificati n of-Motor rated Valves The inspector reviewed an RG&E 10CFR 21 notification, dated May 31, 1990, regarding motor operated valve (MOV) loose screws and sockets for torque switches (TS).
These discrepancies were identified by the licensee on June 1989, during their electrical control configuration drawing upgrade project walkdown of 107 motor operated valves.
The review identified that 26 MOVs were installed with SMA type torque switches.
Of these, 10 MOVs had loose sockets and 9 had loose termination bolts.
Corrective action was taken immediately to tighten the'hardware and to verify the tightness during routine preventive maintenance.
The three EQ valves that had SMA torque switches were MOVs 860A, C and D.
But, only MOV 860A had the loose screws.
The licensee stated that out of the 17 EQ valves
'nstalled at Ginna, only one valve is affected by the Part 21 notification.
The licensee maintenance department evaluations conducted on October 1989 and April 1990, indicated that SMA style torque switches were supplied during 1967 by Limitorque with SMB-00 actuators and loose screw/bolts were due to age/vibration.
The licensee initiated a corrective action report (CAR) 1977 to t'rack this'issue.
It appears that the Part 21 notification was appropriate to identify the loose mounting bolts and termination screw issue, but it was not
.timely.
Further, the licensee failed to perform adequate evaluation and documentation,to establish the environmental qualification for the three EQ valves.
The Part'21 notification did not address the environmental qualification impact on these valves.
According to Limitorque Corporation, these torque switches were not qualified for EQ applications.
The lack of environmental qualification for SMA torque switches was reported to the NRC in April 1988, by the DC Cook Nuclear Power Plant in accordance with 10CFR Part 21
~
Limitorque Corporation also issued a maintenance bulletin in August 1988, to address this issu The licensee stated that the original qualification documentation for these operators was provided by Limitorque and documented in test report B0003,
'ut this report did not identify the type of torque switches installed on these valves.
The licensee also stated that, subsequent to the receipt of the above notifications, RG&E re-reviewed the qualification requirements for all the valves installed with SMA type torque switches in accordance with "Guideline for Evaluating EQ of Class 1E Electrical Equipment in,Operating Reactors" (DOR) guidelines for the required qualified parameter (radiation) and considered them'to be qualified.
This review was neither documented by the licensee nor any justification for continued operation was determined.
The inspector raised concerns regarding the qualification analysis and documentation for these switches.
During this period, RG&E was also in the process of implementing (spring outage of 1989) an MOV refurbishment program due to the vintage of Ginna Station.
In 1989, all EQ MOVs were inspected and the specific locations of the SMAs for the 3 EQ MOVs were identified.
This program included the replacement of these torque switches with qualified parts in March 1990.
At the end of this inspection, the licensee prepared an EQ analysis (EWR No.
4237.35) to determine the EQ impact on Limitorque SMA torque switches prior to replacement during the 1990 outage.
The analysis concluded that the torque switches were fully qualified during their installation and the potential loose mounting screw and termination screw problems did not contribute to the accelerated degradation when exposed to a harsh environment.
Furthermore, EQ performance requirements of MOVs 860A, C&D indicated that since the valves are installed outside the containment and they are required to operate within one minute post LOCA, the radiation dose when they-perform their safety function (to open) would be significantly less than the threshold of SMA torque switch material (laminated phenolic).
The torque switch is bypassed during the initial opening cycle and would perform its safety function.
Also, MOV 806B was available for the operation of one containment spray ring and the components were qualified for harsh environment.
The inspector concluded that generally, the analysis was found to be adequate and meets the requirements of the DOR guidelines.
However, the thoroughness of the analysis could have been improved as evidenced by the following: (1) The analysis never stated the valve actuator type, (2) The analysis addressed only post accident radiation total integrated dose for 30 days and not for the service life of the valve even though the dose is much less than the accident dose, and (3) The analysis did not discuss seismic vibration, even though it is specifically required by 50.4 The absence of documentation to establish environmental qualification for the SMA type torque switches installed in MOVs 860 A, C, and D is a violation of 10CFR 50.49 (50-244/92-01-01).
The timeliness of the submitted RG&E 10CFR Part 21 report regarding loose screws and sockets is considered to be an unresolved item pending further NRC review (50-244/92-01-02).
Also, the timeliness of corrective action taken for unqualified SMA type torque switches as reported in the D.C. Cook Nuclear Power Plant 10CFR Part 21 and Limitorque Maintenance Bulletin identified in April and August of 1988 is considered an unresolved item (50-244/92-01-03).
6.0
'onclusions The inspectors reviewed the RGEcE
'Ginna Nuclear Power Plant engineering organization at the offsite nuclear services engineering facility and at the Ginna Nuclear Power Plant site and found the effectiveness of these organizations to be that commensurate with requirements for the safe and reliable generation of power from nuclear energy.
The engineering organization displayed a cooperative working attitude between onsite and offsite engineers.
Communication between the engineering organizations was good as a result of planned opportunities forjoint participation in plant issues.
An abundance of engineering initiatives to improve engineering performance were evident including drawing improvement, design basis document reconstitution; probabilistic risk assessment, prioritization and configuration management.
Design and facility modifications implemented satisfactorily include those of a major nature, such as the diesel generator building modification and steam generator replacement program.
The inspection identified a violation and two associated unresolved items related to environmental qualification (EQ) of motor operated valves and 10CFR 21 reportability and corrective action.
These issues date back prior to this past year and do not affect an assessment of engineering performance over. the past year.
7.0 Mana ement Meetin s
Licensee management was informed of the scope and purpose of this inspection at the beginning of the inspection at the entrance meeting on January 6, 1992.
The findings of the inspection were discussed with licensee representatives during the course of the inspection and presented to licensee's management at the exit meeting on January 10, 199 ~AA iiMBNTi Pers n
n R
he ter s and Electric o
ration
- R. Baker, Lead Electrical Engineer
- T. A. Daniels, NS&UGinna PRA Project
- R. Davis, QA Engineer-Design
- C. A. Forkell, Jr., Manager, Electrical Engineering
- G. Goetz, Manager, Structural & Construction Engineering
- R. Jaquinn, NS&L
- M. Kennedy, Director, Configuration Management Project
- T. Marlow, Supervisor, Ginna Production
-* R. C. Mecredy, Vice President, Ginna Nuclear Production
- T. Newberry, Lead Mechanical Engineer A. P. Rochino, Lead Mechanical Engineer L. Sucheski, Lead Structural Engineer G. Voci, Manager, Mechanical Engineering
- T. Werner, Technical Assistant, Department Manager J. A. Widay, Plant Manager
- P. Wilkens, Nuclear Engineer, Services Department Manager
- G. Wrobel, Manager, Nuclear Safety & Licensing Nuclear Re ulato nimi i n P. K. Eapen, Chief, System Section, Division of Reactor Safety
- H. Gregg, Senior Reactor Engineer A. Lohmeier, Senior Reactor Engineer
- R. Mathew, Reactor Engineer
- T. Moslak, Senior Resident Inspector Asterisk (*) indicates attendance at exit meetin )
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