IR 05000259/1993004
| ML18036B214 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/17/1993 |
| From: | Branch M, Matt Thomas NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18036B213 | List: |
| References | |
| 50-259-93-04, 50-259-93-4, 50-260-93-04, 50-260-93-4, 50-296-93-04, 50-296-93-4, NUDOCS 9303300041 | |
| Download: ML18036B214 (23) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report Nos.:
50-259/93-04, 50-"60/93-04, and 50-296/93-04 Licensee:
Tennessee Valley Authority 3B Lookout Place 1101 Harket Street Chattanooga, TN 37402-2801 Docket Nos.:
50-259, 50-260, and 50-296 License Nos ~
Browns Ferry 1, 2, and
Inspection Conducted:
February 8-12, and 22-26, 1993 Inspector:
H.
Thomas Date Signed Accompanying Inspectors:
H. Whitener H. Hiller Approved by:
H. Branch, Chse Test Programs Section Engineering Branch Division of Reactor Safety D te igned SUHHARY Scope:
This routine, announced inspection was conducted in the areas of design changes and modifications, engineering support activities, and open item followup.
Results:
Violations or deviations were not identified in the areas inspected.
Desi n Chan e Control The licensee has implemented a process for prioritizing and scheduling plant modifications that address nuclear safety.
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The completed modifications reviewed were considered adequately implemented.
The licensee has established and is implementing a Design Basis Documentation (DBD) program and the means to continually update these documents through the design change control process.
The licensee has adequate controls to ensure that applicable design documents are updated in a timely manner to reflect the as-built plant.
En ineerin Su ort The engineering staff provided adequate
.and timely support to the plant.
The various site engineering organizations were adequately staffed with knowledgeable personnel.
Training provided to engineering personnel was adequate.
guality Assurance (gA) identified one problem with updating training that has since been resolved.
Technical support system engineers demonstrated knowledge and ownership of their systems.
Operations support by both Site Engineering and Technical Support responding to operations issues was considered satisfactory.
The results of probabilistic risk assessment/individual plant examinations were being used by Site Design Engineering.
Self assessment activities were performed by engineering and gA to identify areas for improvement.
0 en Items Inspector Followup Item 50-259/296/93-04-01 was opened to identify the followup of corrective action implementation for the Static "0" Ring Pressure switches for Units I and 3.
Strenqth The good controls and knowledgeable engineering supervision observed for the on-going control room design review (CRDR) field modifications was considered a strengt REPORT DETAILS Persons Contacted Licensee Employees
- R. Albright, Staff Project Engineer
- S. Austin, Licensing Engineer
- H. Bajestani, Technical Support Manager
- R. Baron, Nuclear Assurance and Licensing Manager
. *S. Brown, Support Services Modifications Manager J.
Bynum, Vice President Nuclear Operations L. Clardy, Audit Supervisor, Nuclear Assurance and Licensing
- J. Cory, Radiation Control Manager
- J. Hayes, Modifications Outage Engineering Supervisor
- H. Herrell, Operations Manager H. Jones, Site Engineering Senior Staff Specialist J. King, Modifications Senior Shift Task Manager H. Lawrence, Modifications Manager
- R. Lilleston, Recovery Support Project Engineer
- E. Long, guality Assurance Specialist
- J. Haddox, Site Engineering Manager J.
Ownby, Site Engineering Support Manager
- G. Pierce, Site Licensing Manager
- J. Rupert, Engineering and Modifications Manager
- E. Ridgell, Compliance Manager
"J. Scalice, Plant Manager H. Schwan, Change Control Board Manager F. Tsakeres, Operations Support Manager, Site Engineering J. Wallace, Site Licensing Engineer J. Walton, Design Control Manager, Site Engineering S. Wetzel, Licensing Engineer
- 0. Zeringue, Site Vice President Other licensee employees contacted during this inspection included engineers, operators, craftsmen, and administrative personnel.
NRC Resident Inspectors
- C. Patterson, Senior Resident Inspector
- T. Liu, NRC Intern
- J. Monday, Resident Inspector
- R. Husser, Resident Inspector
- Attended exit meeting Acronyms and initialisms used throughout this report are listed in the last paragrap,
Design Changes and Plant Hodifications (37700)
a
~
Plant Hodifications to Improve Reactor Safety The inspectors reviewed the initiatives taken by the licensee to identify and implement plant modifications to improve reactor safety.
Documentation reviewed during this effort included:
Unit 2 Cycle 6 Hodification Work List Unit 2 Cycle 6 Intermediate Schedule Haster Issues List Business Practice 204, Project Hanagement Business Practice 205, Issue Hanagement The Issue Hanagement Program is the process used for identifying, scoping, evaluating, and approving emergent capital, recovery, and non-routine major 0&H work.
This is also the process by which DCRs are initiated and approved.
An Issue can be initiated by any plant personnel.
All Issues are pre-screened for technical merit by the Technical Support Department to determine if the Issue.
should be included in the HIL data base for further evaluation by the CCB.
The CCB is a committee established by the Site Vice President to screen requested design changes, Issues, and projects.
The CCB reviews all proposed Issues for technical merit and implementation schedule, and makes recommendations to the appropriate authorization level.
The Issues were prioritized and categorized by the CCB according to the nuclear safety significance of the Issues.
The inspectors noted that 460 Issues were initiated in 1992, and at the time of this inspection,
Issues had been initiated in 1993.
The inspectors reviewed selected Issues to verify that they were prioritized and categorized in accordance with Business Practice 205.
Based on discussions with licensee engineering personnel, and review of the above documentation, the inspectors concluded that the licensee has a satisfactory prioritization process for identifying and implementing plant design changes to improve reactor safety..
Violations or deviations were not identified in the areas inspected.
b.
Planning, Development
'and Implementation of Plant Hodifications The inspectors reviewed the DCNs and modification packages listed below to: (1) determine the adequacy of the
CFR 50.59 evaluations performed; (2) verify that the DCNs were reviewed and approved in accordance with TS and applicable administrative controls; (3) verify the subject modifications were installed
(for those that could be physically inspected)
in accordance with the DCN package; (4) verify that applicable plant operating and design documents (drawings, plant procedures, FSAR, TS, etc.)
were revised to reflect subject modifications; (5) verify that the modifications were reviewed and incorporated into the operations training program as applicable; and (6) verify that post modification test requirements were specified and that adequate testing was performed.
In addition, several on going DCN work packages were reviewed and the work was observed.
The following DCNs were reviewed:
DCN H7400A, EECW Strainer Motor Thermal Overloads This modification replaced the obsolete electrical components that were installed in the EECW strainer control cabinets for the strainers motors.
The components included the thermal overload relays, the overload heater elements, and the motor starters.
The original components were produced by the Clark Manufacturing Corporation and have not been made since 1975.
The safety evaluation stated that the existing components caused spurious pump motor trips and are no longer being manufactured:
therefore, equivalent substitutes would be procured from General Electric.
gualified Class 1E motor starters and thermal overload heater elements were installed in the ECCW A, 8, C,
and D strainer electrical control cabinets.
The inspector conducted a walkdown inspection and verified the modifications was satisfactorily completed.
- DCN W17066, RHR Electrical Interlocks Division II This modification added electrical interlocks between RHR shutdown cooling suction MOVs 2-FCV-74-25 and 2-FCV-74-36 and the suppression pool isolation MOV 2-FCV-74-71.
The interlocks were added as the result of INPO Significant Operating Experience Report (SOER), 87-2, Inadvertent Draining Of Reactor Vessel To Suppression Pool At BWRS.
The interlocks were added to prevent the opening of the RHR suppression pool isolation valve when either of the RHR shutdown cooling suction valves were open or vice versa.
These interlocks can prevent the draining of the reactor vessel during shutdown and refueling modes of operation due to improper RHR valve lineup.
This modification to connect the interlock switches required the addition of conduit and wiring between junction boxes JB 2271 and JB 2310 and MOVs 2-FCV-74-25, 36, and 71.
The inspector conducted a walkdown inspection to verify the modification was performed and the Eg requirements were satisfied.
This modification was for Division II.
The licensee also performed the same modification for Division I.
- ECN P0614, Modify The Main Steam Isolation Valves This valve modification implemented the installation of a larger valve stem, enlargement of the bonnet back seat and stuffing box
to accommodate the larger stem and the installation of an anti-rotation device on the valve poppet.
These changes were implemented as part of an extensive HSIV leakage reduction program to valves 2-FCV-01-26, 37, 38, and 52.
These valves had repeatedly exhibited excessive leakage when tested in accordance with the Appendix J leak rate test and TS requirements.
The safety assessment considered seismic conditions, reactor pressure boundary integrity, valve operating requirements, and the flow rate and pressure drop.
Additionally, vendor concurrence was obtained prior to implementing the modification.
Post modification functional testing was performed as part of system operability verification prior to restart of Unit 2 and as part of the partial ECN close out.
The inspectors concluded that the design change package was adequate.
- DCN W-20321A Cut, Cap, and Hydrostatic Test EECW Flush Lines This modification removed the 1.5 inch flush lines connected to the 1D EDG engine cooler North and South EECW supply lines.
The two flush lines tied together through check valves 0-67-504 and 0-67-505 into a
common line.
The common line tied into a drain line from the standby gas treatment system drain sump pump through isolation valve 0-67-506.
The drain line discharged into the radwaste system.
The DCN provided for cutting the flush lines at the connections of the North EECW supply line, the South EECW supply line, and the radwaste drain line.
The cuts at the EECW lines and the radwaste drain line were capped and welded to preserve the integrity of the system boundaries.
The lines original purpose was to flush EECW water to the radwaste system.
However, the lines were no longer used.
The removal of these lines reduced the potential for leaks.
The safety consideration for removal of these lines was the preservation of the EECW and radwaste boundary integrity.
Hydrostatic testing was performed to verify system tight integrity.
The inspector reviewed the
CFR 50.59 safety evaluation and determined that the evaluation bounded the scope of the DCN.
The evaluation addressed the impact of the modification on all affected systems; identified required document updates including DBD; and included appropriate design inputs such as fire protection, pipe breaks, seismic conditions and pipe vibrations, Additionally, post-modification testing was specified in the DCN which required a hydrostatic test on all new welds and associated piping.
The inspectors concluded that this DCN met the design change control process in effect at Browns Ferry.
This DCN is an example of timely Site Engineering support to plant operations.
The modification was implemented to correct EECW
leakage through the corroded flush lines.
Since the leakage had the potential to affect operability of the 1D EDG, the plant =was required to enter a seven day limiting condition of operation according to TS 3.9.P.3.
The modification package was developed and implemented within four days, DCN H-7844A, Diesel Generator Tank Relief Valves This DCN involved removing four of the five relief valves originally installed on each bank of the Units I/2 EDG starting air systems.
The inspectors reviewed the
CFR 50.59 safety evaluation and verified that the safety evaluation adequately bounded the scope of the DCN.
This included such design inputs as pipe vibration, seismic, and the proper setpoint and relief capacity of the remaining relief valve for each bank.
The inspectors also performed a field walkdown of the Units 1 and
EDG rooms and verified that field installation was in accordance with the DCN.
The inspectors also performed a walkdown of the Unit 3 EDG rooms and observed that the same modification had also been implemented for Unit 3.
The inspectors determined from the information reviewed that this DCN was implemented in accordance with licensee procedures and controls.
- DCN W-16945A, Repair Residual Heat Removal Service Water Discharge Piping This DCN installed a welded flange with a bolted on blind flange in one of the RHRSW system discharge piping lines.
This modification was required to isolate the piping because a pipe failure had allowed warm RHRSW discharge water to flow into the cool water channel instead of the river when Unit
RHRSW pumps were operating.
The inspectors reviewed the
CFR 50.59 safety evaluation to verify that it bounded the scope of the DCN.
This included consideration of such design inputs as seismic, material compatibility, equipment failure modes, pipe stress, and decay heat removal capability.
The inspectors determined from the information reviewed that this DCN was satisfactorily implemented.
Violations or deviations were not identified in the areas inspected.
C.
Observation 'of Modifications in Progress The inspectors conducted walkdown inspections in the control room to examine modification work in progress.
DCNs and work packages were reviewed to ensure all documentation was approved and contained the appropriate work instructions and drawings.
The completed work was compared against the work instructions and drawings to ensure the work requirements were followed.
In
addition, discussions were held with the craftsmen to verify that they understood the work packages.
All the work packages observed required that control panel switches be relocated.
The work included removing the switches and existing wiring, relocating the switches, rewiring the switches, repairing the holes in the panel, and performing the post modification tests.
The inspectors determined the craftsmen and their engineering supervisors were knowledgeable and performed the modifications work in a professional manner.
Portions of the following modifications and work packages were observed:
DCN W19557A Work Package WP 2823-92, Core Spray Cooling Panel 2-9-3 This work package included the following components:
2-HS-75-9A) 2-HS-75-14A, 2-HS-75-23A, 2-HS-75-25A
DCN W19557A Work Package WP 2822-92, Core Spray Cooling'Panel 2-9-3 This work package included the following components:
2-HS-75-2A, 2-HS-75-5A, 2-HS-75-11A, 2-HS-75-22A, 2-HS-75-59)
2-XS-75-61
DCN W19563A Work Package WP 2988-92, Hain Steam Panel 2-9-3 This work package included the following components:
2-HS-01-15AKB, 2-HS-01-27AKB, 2-HS-01-38B)
2-HS-Ol-52B)
2-HS-01-58A, 2-HS-01-59A
DCN W19558A Work Package WP 2909-92, Residual Heat Removal Panel 2-9-3 This work package included the following components:
2-HS-74-103)
2-HS-74-120)
2-HS-74-98A)
2-HS-74-149, 2-HS-23-35A)
2-HS-23-39A
DCN W17364A Work Package WP 4621-92, Diesel Generator
"C" Panel Standby DG System 4-KV Shutdown Board This work package included the following components:
O-HS-211-CD2, O-HS-43-211-CD2, 0-HS-82-OC/1A 0-HS-82-OC/2A, 0-HS-82-OC/3A, 0-HS-82-OC/4 Violations or deviations were not identified in the areas inspecte d.
Configuration Control The inspectors reviewed the licensee's program and procedures that were developed and implemented to maintain configuration control of the applicable design documents and drawings.
The program was examined to ensure that design control was maintained and that documentation affected by DCNs was updated in a timely manner to reflect the as-built plant.
The documents reviewed for the configuration control program included the following procedures and Site Standard Practices (SSP):
- SSP-2.8 Drawing Control
- SSP-2. 11 Drawing Deviation Program
- SSP-9.3 Plant Hodification And Design Change Control
- BFEP PI 89-06 Design Change Control These procedures describe the controls for correcting and updating applicable plant documents as a result of a DCN, drawing deviation, or a plant modification.
The inspectors reviewed these procedures to verify that their requirements were implemented for the DCNs discussed in paragraph 2.b.
SSP-2.8, Drawing Control, this procedure establishes the process for maintaining and controlling drawings, including their distribution by the Documentation Control Group. Its main purpose is to ensure
"controlled" drawings are used.
SSP-2. 11, Drawing Deviation Program, this procedure establishes the requirements for dispositioning discrepancies between the actual plant configuration and the as-constructed drawings or configuration control drawings.
SSP-9,3, Plant Modification And Design Change Control, this procedure establishes the administrative controls and requirements for design changes and for plant modifications based on design changes.
The scope of this procedure covers the design, installation, return to operation, and closeout of design changes except for fuel design and temporary alterations.
The inspectors reviewed SSP-9.3 and determined that it contained all the necessary informatioh and requirements to ensure that plant configuration would be maintained after modifications.
BFEP PI 89-06, Design Change Control procedure, this engineering procedure provides the requirements for implementing the design change control process and maintaining configuration control using the Design Change Notice (DCN).
Configuration control for engineering includes revising and maintaining the plant drawings to reflect the as installed condition of the plant.
However, in PI 89-06, Section 21, Issue/Revisions To Drawings Without DCNs,
administrative revisions to plant drawings without using a
DCN are allowed.
The administrative revisions are to be nontechnical in nature and for documentation purposes only.
The resident inspectors have identified a concern that questionable administrative revisions have been incorporated into several drawings.
This concern is discussed in the NRC Resident's Report 50-259, 50-260, and 50-296/93-02.
The inspectors determined that the licensee has implemented an acceptable program to maintain configuration control for plant.
modifications.
The procedures reviewed contained the necessary requirements to develop DCNs and for the implementation of plant work packages.
Violations or deviations were not identified in the areas.inspected.
e.
Haintenance of the Design Basis Document Through the DCN Process Design Basis Documentation (DBD) consists of Design Criteria and Configuration Control Drawings.
Prior to restart of Unit 2 the licensee completed a Design Basis Verification Program.
This program was a one-time verification that Unit 2 design criteria were current and complete.
The inspector briefly reviewed selected design criteria and determined that the Design Criteria are written documents which specify the design requirements for both systems and components of the systems.
The licensee also developed Design Criteria to support Unit 3 restart.
They consist of applicable Unit 2 criteria with additional requirements unique to Unit 3 and Units 2 and 3 interfaces.
The licensee indicated that Design Criteria are developed for about 45 systems including safety systems and some non-safety systems important to safety.
An additional part of the DBD are the Constructed/Configuration Control Drawings.
These drawings represent the current as constructed condition of the plant and are discussed in detail elsewhere in this report, Hethods of maintaining Design Basis Documents current were discussed with Site Engineering.
If a Design Change Notice (DCN) impacts the design criteria, a Design Information Hemorandum (DIH) would be generated and packaged with the affected design criteria.
DIHs are periodically incorporated into the Design Criteria. If a "configuration control drawing" considered important to plant safety and operation was changed by a plant modification or design change, the drawing would be updated prior to return of the affected system to service.
The design change control closure process controls these actions and requires all affected documentation to be updated prior to final DCN close out.
Although there was no intent to review the design criteria for technical adequacy during this inspection, the inspector did determine that the licensee has established design basis documents
for plant systems and the means'o continually update these documents through the design change control process.
Violations or deviations were not identified in the areas inspected.
I'.
PRA Application in Design Control Discussions with licensee personnel revealed that the licensee completed an Individual Plant Examination (IPE) for Browns Ferry Unit 2 and submitted it to the NRC in September 1992.
Licensee personnel indicated that, Site Engineering and other plait groups were involved in reviewing the PRA during development to ensure that the PRA was technically and administratively correct.
The PRA is considered in the design change process.
Procedure PI-89-06, attachment 10, addresses reviewing DCNs for impact on the PRA.
Licensee personnel also indicated that the PRA will be updated after the Unit 2 outage to reflect DCNs implemented since December 1991.
The inspectors concluded that the PRA is another tool that is'ow available to the design organization which should enhance the design control process.
Violations or deviations were not identified in the areas inspected.
Engineering and Technical Support Activities The inspectors reviewed activities performed by the various engineering and technical support groups in an effort to assess the effectiveness of the support being provided to the plant.
The inspectors concluded that effective engineering and technical support was being provided.
a
Problem Identification and Resolution The inspectors noted that licensee engineers identified and resolved technical problems for both operations and maintenance.
Problems were identified by and assigned to the Operations Support Group in Site Engineering (design)
and the plant Technical Support Department (systems engineering).
Problems were identified through various means which included daily plan of the day (POD)
meetings, work orders, plant issues, engineering evaluation requests, DCN implementation, plant walkdowns, verbal requests, etc, Site Engineering Operations Support and Technical. Support engineering personnel provided adequate and timely responses to the items that were assigned to them.
The inspectors reviewed selected items that were assigned to either Site Engineering Operations Support or Technical Support.
These examples of real time support through problem identification and resolution included problems identified in the plant Issue process and the following:
This DCN was prepared and implemented in order to remove
Unit 2 from a LCO.
Thi s DCN i s di scussed in greater detail in paragraph 2.b. of this inspection report.
- DCN W-21852, involved cutting and capping unnecessary vent and drain lines in the recirculation and RHR systems in order to prevent additional outage impacts during plant operations.
- DCN W-21976, involved installing a portable total organic carbons filter in the main turbine lube oil purifier piping connection in order to remove particulate to the turbine journal bearings and keep other turbine internal components clean to reduce the potential for a turbine trip.
- Assistance provided by Technical Support to resolve a traversing incore probe (TIP) problem.
- System engineer provided support to operations during a fall 1992 Unit 2 startup when a control rod would not withdraw from position 00.
The recommended actions provided by the system engineer were performed by operations and resulted in the control rod being successfully withdrawn.
- The system engineer identified a condition during the performance of a routine walkdown of the raw cooling water system in December 1992, that did not allow a portion of the secondary containment atmosphere to be filtered through the standby, gas treatment system.
This condition resulted in licensee event report (LER) 50-259/92-005, dated January 12, 1993.
The inspectors also reviewed documentation where Technical Support personnel were commended by plant management for other instances of support to the plant.
The inspectors determined from the information reviewed that timely and satisfactory engineering support has been provided by the applicable engineering organizations.
Licensee engineers initiate actions based on emerging technical issues.
Violations or deviations were not identified in the areas inspected.
b.
Organization, Staffing, and Training The inspectors reviewed the licensee's organization and staffing to determine whether the licensee's engineering and technical support organizations were adequately stafFed to provide effective support to the plant.
The Engineering and Modifications organization is comprised of Site Engineering (design)
and Site Modifications (field implementation).
The Site Engineering department has a staff of 171 personnel, which also includes the Engineering Recovery Group for Unit 3 and the Operations Support Group.
The Operations
Support Group was formed in 1991 to provide design interface and real time discipline design support to the operating plant.
Site Hodifications has a staff of 39 personnel.
Site Engineering and Site Hodifications are both supplemented by contract personnel during outages and other periods of increased work load.
The Technical Support staff consists of 94 personnel which includes Reactor Engineering and the Restart Test Program (RTP)
group for Unit 3 restart.
Technical Support is also supplemented by contract personnel.
The inspectors noted that Technical Support systems engineers demonstrated knowledge and ownership of their assigned systems and provided support for maintenance, operations, and assigned surveillance procedures.
The inspectors determined that the engineering and technical support staffing levels were adequate to provide effective support to the plant as evidenced by the completed DCNs reviewed, the examples discussed in paragraph 3.a.
above, and other performance indicators reviewed by the inspectors.
Examples of performance indicators included:
1)
154 of the 160 DCNs initially identified in June 1991 for the Unit 2 outage, plus approximately
additional DCNs identified after June 1991 were issued by the July 1992, target date of six months prior to the= outage, 2) completed 1487 drawing deviation only DCNs during fiscal year 1992, which exceeded the goal of 1200; 3) reviewed all plant Issues for technical merit prior to submittal to the CCB; 4) six of six incident investigations were completed on time by Technical Support; and 5) eight of 'eight temporary alteration control forms (TACF) were properly prepared by Technical Support.
In addition to reviewing the licensee's organization and staffing, the inspectors also reviewed the training provided to the engineering and technical support staffs.
Training requirements for Site Engineering personnel are provided in Standard Engineering Procedure SEP-9. 1.2, Training of Personnel, and the Nuclear Engineering Training Requirements Matrix.
The training matrix includes training requirements for managers and discipline specific training.
Training requirements for Technical Support personnel are described in procedure TSBI-007, Technical Support Engineer Training Requirements.
In addition,
CFR 50.59 Safety Evaluation initial and continuing training is provided in accordance with procedure EGT024.007, gualified 50.59 Preparer Training.
The licensee maintains a list of qualified 50.59 safety evaluation preparers and reviewers.
The inspectors reviewed the list of qualified safety evaluation preparers and reviewers and verified that the individuals who prepared the 50.59 safety evaluations for the DCNs discussed in paragraph 2.b. of this inspection report were on the lis During further review of engineering training the inspectors noted that during QA Audit BFA92304 dated December 29, 1992, site QA identified that Site Engineering personnel had not completed required reading training for selected procedures.
Significant Corrective Action Report (SCAR)
BFSCA920024304 was initiated to address the recurring problem.
Site Engineering had responded to the SCAR and was in the process of implementing corrective actions to address the finding.
The inspectors concluded that, except for the QA finding, the training provided to engineering and technical support personnel was considered adequate.
Violations or 'deviations were not identified in the areas inspected.
Self Assessment The inspectors discussed self assessment methods with licensee personnel and found that the licensee has an active self assessment program using multiple techniques.
Principal in the assessment process was the Site QA organization.
QA was involved in the performance of detailed
- assessments in selected areas, as well as auditing and monitoring both TVA and contractor procedures, programs, and performance.
Additionally, QA performs an overview function such as review of the site engineering technical assessment of the design Change Control Program and Bechtel's QA Surveillance on Calculations.
Site QA also generates the Nuclear
"
Power Nuclear Assurance Level 1 Trend Analysis Report on a quarterly and annual basis.
This report provided a brief description of the problems and the trends for the seven SALP areas.
Review of the 1992 annual and first quarter of 1993 trends showed that QA has rated the Nuclear Engineering and Technical Support areas as satisfactory.
Site Engineering has developed twelve performance indicators to track-engineering performance in the design change and closure process.
The rating on performance indicators was issued monthly and trended on a
quarterly basis to assess the performance of Site Engineering.
Three
.indicators (1)
As Built Drawings (2)
DCN Closures (3)
D-DCN Design Basis=
Reviews were classified as "Poor", since the time to implement them exceeded the targeted goals.
The inspectors did not consider the "Poor" rating to be an unsatisfactory condition since the licensee met the requirements in their procedures.
Three indicators (1)
PORC Rejections (2)
DCNS Cost Variance (3)
F-DCNs Due To Engineering Error were classified as "Outstanding".
Technical Support has also developed quality indicators to assess the quality of the support they were providing to the plant.
The inspectors reviewed the monthly quality indicator reports for the past eight months and noted that, although there were instances where the need for improvement was identified, the overall quality of the support provided by Technical Support was considered satisfactor The inspectors considered the licensee's use of the various performance indicators as acceptable methods for self assessment and demonstrated the licensee's commitment to improving the quality of the engineering and technical support being provided to the plant.
Violations or deviations were not identified in the areas inspected.
Followup Open Items (92701)
(closed) Bulletin IEB 86-02, Static "0" Ring Differential Pressure Switches (for Units 1,2',
and 3).
IEB 86-02 was issued by the NRC in July 1986, as a result of problems identified with the Static "0" Ring Series 102 or 103 differential pressure switches.
A number of events occurred where the switches did not actuate within the set point tolerance or failed to actuate.
Testing by some licensees indicated that switches performed erratically.
Some of the problems identified were failure to actuate due to corrosion, shift in set point, offset due to calibration, and sensitivity due to exposure to operating conditions.
The licensee's response to IEB 86-02 dated July 20, 1987, indicated that new switches were to be install,ed in each unit prior to restart as part of the environmental qualification upgrade program.
The switches were to be used to control RHR minimum flow valves 2-FS-74-50 and 2-FS-74-64.
The valves close to isolate the minimum flow lines on an injection to obtain maximum flow to the reactor pressure vessel and open the minimum flow recirculation lines on low flow conditions to provide adequate cooling to pump seals and bearings.
The licensee also subsequently installed these same type of switches to control the core spray minimum flow valves 2-FS-75-21 and 2-FS-75-49.
This issue was closed for Unit 2 prior to restart.
After reviewing the status of this issue for Units 1 and 3, the inspector determined the licensee met the intent of providing corrective action for IEB 86-02.
Bulletin IEB 86-02 is closed for Units
and 3.
An Inspector Followup Item (IFI) 50-259/50-296/93-04-01, Install Static "0" Ring Switches For IEB 86-02 Prior To Restart was opened for Units 1 and 3 since implementation has not been initiated.
This IFI is a "restart item" for Units 1 and 3.
(open)
IFI 50-259/50-296/93-04-01, Review Static "0" Ring Switches Per IEB 86-02 Prior To Restart.
Exit Interview The inspection scope and results were summarized on February 26, 1993, with those persons indicated in paragraph 1.
The inspectors described the areas inspected and discussed in detail the inspection results.
Proprietary information is not contained in this report.
Dissenting comments were not received from the license Acronyms and Initialisms AFW BFEP BWR CCB CRDR CTS DBD DCN DCR EDG EDSFI EECW EOP EQ FCV FS FSAR GL HS IEB IFI INPO KV MIL MOV MSIV NEP NRC 0&M PI PORC PRA QA RHR RHRSW SALP SSP SW TS TVA Auxi1 i ary Feedwater Browns Ferry Engineering Procedure Boiling Water'eactor Change Control Board Control Room Design Review Commitment Tracking System Design Basis Documentation Design Change Notice Design Change Request Emergency Diesel Generator Electrical Distribution System Functional Inspection Emergency Equipment Cooling Water Emergency Operating Procedure Environmental Qualification Flow Control Valve Flow Switch Final Safety Analysis Report Generic Letter Hand Switch Inspection Enforcement Bulletin Inspector Followup Item Institute of Nuclear Power Operations Kilovolt Master Issues List Motor Operated Valve Main Stream Isolation Valve Nuclear Engineering Procedure Nuclear Regulatory Commission Operations and Maintenance Project Instruction Plant Operations Review Committee Probabilistic Risk Assessment Quality Assurance Residual Heat Removal Residual Heat Removal Service Water Systematic Assessment of Licensee Performance Site Standard Practice (Procedure)
Service Water Technical Specifications Tennessee Valley Authority