ML20057G183
| ML20057G183 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 10/15/1993 |
| From: | Johnson W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20057G182 | List: |
| References | |
| 50-498-93-31, 50-499-93-31, NUDOCS 9310210005 | |
| Download: ML20057G183 (34) | |
See also: IR 05000498/1993031
Text
{{#Wiki_filter:. _ _- - - . __ _ . . . , v. . j - j i 'M . ] APPENDIX .i U.S. NUCLEAR REGULATORY COMMISSION -! REGION IV , i NRC Inspection Report Nos.: 50-498/93-31 l 50-499/93-31 l Licenses: NPF-76 NPF-80 j Licensee: Houston Lighting & Power Company 7 P 0. Box 1700 Houston, Texas 77251 1 i Facility Name: South Texas Project Electric Generating Station (STPEGS), l Units 1 and 2 Inspection At: Region IV Offices, Arlington, Texas . ! Inspection Conducted: September 20-24, 1993 l > Inspectors: Mark A. Satorius, Project Engineer, Project Section A,-Division of l Reactor Projects l T. O. McKernon, Reactor Inspector, Operations Section, Division of.
Reactor Safety
i I Approved: [A// /O/f/93 ' _ '; W.D.Jopson,Chie~f,ProjectSectionA Date ' ~ t Inspection Summary
1 Areas Inspected: Routine in-office inspection of the issues' contained in the
Diagnostic Evaluation Team (DET) Report, Confirmatory Action Letter (CAL) and Supplements, the licensee's Operational Readir.ess Plan (ORP), routine and
special NRC inspection reports, licensing acnions, and NRC staff actions. ' Results: 1
No violations or deviations of NRC requirements were identified. l
The DET report, CAL and Supplements, ORP, routine and special NRC j
inspection reports, licensing issues, and NRC staff actions assigned by the NRC Executive Director for Operations following the Diagnostic I Evaluation were reviewed. Based on this review, issues that the NRC considers necessary to be addressed prior to the restart of either unit (Restart Issues) were identified and listed in Attachment 2. l t I i ! ~ 9310210005 931015 , PDR ADOCK 05000498 G PDR- l ! ' . - ,
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l 6 i -2- - . L Items identified in the review of the DET report, ORP, and NRC staff e actions related to Restart Issues were assigned an Inspection Followup . I Item (IFI) in order to facilitate tracking and eventual closure. In addition to these items, previously identified NRC inspection items and - L licensing issues that were related to the Restart Issues (e.g., IFIs, unresolved items [URIs], violations, and others) were identified. All of these items related to the Restart Issues were cross-referenced and t incorporated into a matrix in Attachment 2. - Attachment 3 contains a matrix that cross-references items similar in i
' root cause and required corrective action. - ! . Summary of Inspection Findings:
i See Attachments 2 and 3 3 e Attachments: f ! Attachment I - Persons Contacted and Exit Meeting i
Attachment 2 - Restart Issues /Related Items Matrix ! e Attachment 3 - Summary of Inspection Findings / Common Items Reference [
Matrix i ! , ) ' . I . t ! l ! l !
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-3- 7 ! ! DETAILS l [ l BACKGROUND _ P Both units at STPEGS were shut down in early February 1993 and remain shutdown as a result of numerous broad scope problems identified by the NRC and the , licensee. On February 3,1993, following a reactor trip, the Unit 2 turbine-driven ! auxiliary feedwater pump started and immediately tripped on overspeed. On ! February 4, 1993, Unit I was required to shut down as a result of repeated failures of the turbine-driven auxiliary feedwater pump to start on demand and operate without tripping on overspeed. As a result of these turbine-driven auxiliary feedwater pump problems, NRC issued a CAL to the Houston Lighting & .; Power Company on February 5, 1993, and dispatched an augmented inspection j team (AIT) to investigate the details surrounding the-turbine-driven auxiliary
feedwater pump problems. The CAL and Supplement, which was subsequently issued on May 7,1993, identified a number of issues that required resolution prior to either unit being restarted. A second supplement to the CAL was ! issued on October 15, 1993, and identified additional restart issues. In addition to the AIT activities, several special inspections were conducted
since February 1993, in order to resolve safety and regulatory issues identified at STPEGS. Several of these special inspections resulted in enforcement action being taken against the licensee. ! Separate from these turbine-driven auxiliary feedwater pump and other
' problems, the NRC Office for Analysis and Evaluation of Operational Data ! conducted a Diagnostic Evaluation of STPEGS during the period March 29 to i ' April 30, 1993. The findings of this evaluation were forwarded to the - ' licensee on June 10, 1993. Numerous items were documented in this report, including a number of issues that NRC considered of sufficient scope and ! safety significance to require resolution prior to either unit being j restarted. l In initial response to the DET report, the licensee submitted their ORP or, ' August 28, 1993. In addition to responding to short-term problems that the i licensee considered necessary to resolve prior to restart, the ORP addresses { the planned actions in response to the CAL and CAL Supplement of May 7,1993, ! special and routine Regional inspections, and other licensee-identified concerns and problems. j in an effort to identify the issues that NRC considers necessary to address )
prior to restart (Restart Issues), a review was conducted of the DET report,
CAL and CAL Supplements, ORP, routine and special NRC reports, licensing issues, and NRC staff actions. As a result of this review, the Restart Issues i
in the following table were identified. l l l
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RESTART ISSUES l 1 Turbine-Driven Auxiliary Feedwater Pump Reliability and Testing l ' Methodology 2 Station Problem Report Process, Threshold, Licensee's Review of i , Existing Reports for Issues Affecting Operability and Safe Plant ~ - Operation , 3 Service Request (SR) Backlog, Including Reduction Accomplished During the Current Outages and the Licensee's Review of Outstanding i SRs for Issues Affecting Equipment Operability, Safe Plant
Operation, and Operator Work-Arounds l . 4 The Postmaintenance Test Program, including Corrective Actions in .' Response to Violations and Other Process Improvements and the Basis For Licensee's Confidence That Equipment Removed from Service for i i Maintenance is Properly Restored to an Operable Status 5 The Outstanding Design Modifications, Temporary Modifications, and
other Engineering Backlog Items, Including the Licensee's Review of These for Issues Affecting Equipment Operability, Safe Plant i Operation, and Operator Work-Arounds . ! 6 Adequacy of Operations Staffing 7 Adequacy of Fire Brigade Leader Training and Qualifications
8 Adequacy of Fire Protection Computers and Software, the Licensee's l ' Success in Reducing the Number of Spurious Fire Protection System Alarms, and Other Fire Protection Hardware Problems f 9 Licensee Management's Effectiveness in Identifying, Pursuing, and l , Correcting Plant Problems ! 10 NRC Review of the Effectiveness of the Licensee's SPEAKOUT Program j 11 Standby Diesel Generator Reliability 12 Essential Chiller Reliability , 4 t l 13 Monitoring of the Licensee's System Certification Program
14 Adequacy of the Licensee's Resolution of the Reliability and e Operability of the Feedwater Isolation Bypass Valves i 15 Tornado Damper Issues
16 Emergency Preparedness Accountability Drills These Restart Issues are listed in Attachment 2 and are referenced to items . related to restart, such as IFIs, URIs, violations, and others. Closure of each of these items related to restart is not necessary for the associated
Restart Issue to be considered resolved. ' , t - ,r , ., -, - n.,- ,- ., , - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
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1 . l ' -5- i ! ! 2 DIAGNOSTIC EVALUATION TEAM ITEMS RELATED TO RESTART (92701) This section was structured to address the issues in the DET Report in a line- j by-line format. The Executive Summary, Section 1.0, " Introduction"; , Section 3.0, " Root Causes"; and Section 4.0, " Exit Meeting" were not addressed i in this line-by-line format because the items addressed in these sections were . ' determined to be identified in the detailed sections of the DET Report or did- a not contain issues associated to unit restart. Similarly, with several noted ' > ' exceptions, the introductory sections of Section 2.1, " Operation"; Section
2.2, " Maintenance and Testing"; Section 2.3, "Engineerir.g Support"; and
Section 2.4, " Management and Organization," were not addressed because the issues addressed in these sections were also determined to be identified in i the detailed portion of the corresponding section of the report. In addition, l the positive observations in Sections 2.1, 2.2, 2.3, and 2.4 were not i addressed because these issues were determined to be not applicable. j ! 2.1 IFIs Identified in the Operations Section of the DET Inspection
' 1 2.1.1 Paragraph 2.1.1, " Marginal Staffing for Scope of Responsibility" f , f 2.1.1.1 (0 pen) IFI 498;499/9331-01: The team found that the assigned ! workload and poor site support adversely impacted the capability of the shift i supervisor and the control room staff to safely operate the plant (identified
in paragraph 2.1). ' , , 2.1.1.2 (0 pen) IFI 498;499/9331-02: Operators were significantly affected by
degraded plant equipment, including equipment workarounds and the ! administrative burden associated with the high rate of removal and return of equipment to service (identified in paragraph 2.1). l t 2.1.1.3 (0 pen) IFI 498;499/9331-03: The shift supervisors and their control
room staff could not effectively maintain the proper focus and overview of ! ! plant operations because of their participation in administrative programs and resource-intensive surveillances.
f 2.1.2 Paragraph 2.1.2, " Poor Support to Operations" l 2.1.2.1 (0 pen) IFI 498;499/9331-04: Poor support to operations was adversely l impacting the licensee's capability to safely operate the plant. l l . 2.1.3 Paragraph 2.1.3. " Confusing and Conflicting Management Expectations , . ! 2.1.3.1 (0 pen) IFI 498;499/9331-05: Management has sent confusing and i conflicting guidance to the control room staff through numerous memoranda j 2 i without soliciting input from the first line supervisors. j 2.1.4 Paragraph 2.1.4, " Inconsistent Operator Performance" No IFls related to Restart issues were identified in this paragraph.
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l -6- .j j 2.1.5 Paragraph 2.1.5, " Ineffective Problem Identification and Resolution" l 2.1.5.1 (0 pen) IFI 498;499/9331-06: Management support to correct program j ' and component problems was not always effective. l
i 2.2 IFIs Identified in the Maintenance and Testina Section of the DET l Inspection , 2.2.1 Paragraph 2.2.1, " Ineffective Corrective Maintenance" l l 2.2.1.1 (0 pen) IFI 498;499/9331-07: The team found that maintenance and ! testing weaknesses reduced the reliability of safety-related and balance-of- l plant equipment (identified in paragraph 2.2). l I ! . 2.2.1.2 (0 pen) IFI 498;499/9331-08: Ineffective corrective and weak preventive maintenance significantly contributed to poor equipment performance
' l (identified in paragraph 2.2). 2.2.1.3 (0 pen) IFl 498;499/9331-09: Ineffective corrective maintenance, caused by inadequate root cause analysis, poor prioritization of work, and
' , poor craft performance, adversely affected safety-related equipment
performance (identified in paragraph 2.2). ' ., 2.2.1.4 (0 pen) IFI 498;499/9331-10: Surveillance and postmaintenance testing I did not always verify equipment operability (identified in paragraph 2.2). -l t 2.2.1.5 (0 pen) IFI 498;499/9331-11: Standby diesel generator (SDG) injector l pump hold down studs failed on nine separate occasions. The root cause ? l analysis was shallow and corrective actions were insufficient to preclude recurrence. The licensee did not perform a more detailed analysis of the stud - failures until the team became involved. i l 2.2.2 Paragraph 2.2.2, "Less than Fully Effective Preventative Maintenance r Program" No IFIs related to Restart Issues were identified in this paragraph. l , i 2.2.3 Paragraph 2.2.3, " Maintenance Training Deficiencies" No IFis related to Restart Issues were identified in this paragraph. l . ! 2.2.4 Paragraph 2.2.4, " Deficiencies in the Replacement Parts Program" t ! No IFis related to Restart issues were identified in this paragraph. l
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_ . - . . . . -. __- . . . , . I . i - 1 -7- , 2.2.5 Paragraph 2.2.5, " Insufficient Support-to Maintenance" , < 2.2.5.1 (0 pen) IFI 498;499/9331-80: Management support to maintenance was ! poor, reducing the effectiveness of the maintenance process and quality of the i maintenance effort. , j 2.2.6 Paragraph 2.2.6, " Inefficient Work Control Process" i , 2.2.6.1 (0 pen) IFI 498;499/9331-12: Several SDG failures resulted from i broken fuel oil injector pump hold down studs, many of which were installed l ' using a deficient stud driver tool designed by the system engineer. The , system engineer failed to consult design engineering or the SDG vendor while j designing the tool. ! ' ) 2.2.6.2 (0 pen) IFI 498;499/9331-79: Work procedures occasionally contained unneeded information and did not match the experience of the individuals using
the procedures. Procedures were sometimes ignored and often revised to ' ' correct errors. i , 2.2.7 Paragraph 2.2.7, " Post-Maintenance Testing Program Not Always j Effective" ! .t . 2 2.?. 7.1 (0 pen) IFI 498;499/9331-13: Nun,erous weaknesses in the
implementation and programmatic requiremonts for postmaintenance testing . reduced assurance that equipment was operable upon return to service. ! t - ' j' 2.2.7.2 (0 pen) IFI 498;499/9331-14: The Post-Maintenance Testing manual used by planners to select the appropriate test requirements did not specify ' appropriate detail and occasionally specified the wrong test. l 2.2.8 Paragraph 2.2.8, " Periodic Testing Not Always Effective" l t " 2.2.8.1 (0 pen) IFI 498;499/9331-15: Previous licensee event reports and NRC l enforcement actions documented that the licensee's testing procedures did not ' ensure all Technical Specification surveillance requirements were being met. j Numerous instances had been identified where procedures were inadequate to , meet Technical Specification surveillance requirements, thereby reducing
assurance that the equipment was operable. Among these was a failure to l ,
completely test a manual reactor trip handswitch and the nonconservative
setting of one of the four reactor protection channels during a reactor ' startup. To address these inadequacies, the licensee committed to perform a J sample review of Technical Specification surveillance tests and verify their j technical adequacy. The licensee's sample indicated that the Technical ! i Specification surveillance program needed strengthening but did appear to
satisfy Technical Specification. The licensee later committed to enhance the
Technical Specification surveillance procedures. ! , ! . ! ' 2 1 I i _ . _ _ _ _ . - _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ - _ _ . _ _ _ , _ - - _ - - - . - - . - -
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. t -8- i i 2.3 IFis Identified in the Engin'aring Support Section of the DET Inspection 2.3.1 Paragraph 2.3.1, " Weak Support in Resolving Plant Problems" 2.3.1.1 (0 pen) IFI 498;499/9331-16: _ Configuration control weaknesses ! adversely affected safety-related equipment and the quality of design i . documents (identified in paragraph 2.3). , 2.3.1.2 (0 pen) IFI 498;499/9331-17: The licensee also did not resolve , several chronic fire protection issues in a timely manner. The issues
included excessive shrinkage of penetration seals, an unreliable fire alarm ! system, a large backlog of service requests on fire' protectio' systems, and j inadequate control of transient combustibles in the plant (ideritified in j paragraph 2.3). 2.3.1.3 (0 pen) IFI 498;499/9331-18: The engineering departments gave weak support in resolving plant problems. The root cause analyses and resulting corrective actions were often ineffective in preventing repetitive equipment problems. , 2.3.1.4 (0 pen) IFI 498;499/9331-77: Torque measurements and computations associated with testing of motor-operated valves (MOVs) were not evaluated to verify valve operability. Other MOV operability / reliability issues existed. 2.3.2 Paragraph 2.3.2, " System Engineering Program Not Effectively , Impiemented"
No IFIs related to Restart Issues were identified in this paragraph. 2.3.3 Paragraph 2.3.3, " Engineering Work Backlogs Were Large, Poorly -Tracked, ) and Not Well Managed"
) ' 2.3.3.1 (0 pen) IFI 498;499/9331-81: Engineering backlogs were large, poorly
tracked, and not well managed. Informational data bases were often inaccurate or not current. -{ 7 2.3.4 Paragraph 2.3.4, "Use of Industry and Site Operational Experience Was j ' Inadequate" No IFIs related to Restart Issues were identified in this paragraph. 2.3.5 Paragraph 2.3.5, " Insufficient Support to Engineering" - No IFIs related to Restart Issues were identified in this paragraph. l ' 2.3.6 Paragraph 2.3.6, " Configuration Control Weaknesses" ! 2.3.6.1 (0 pen) IFI 498;499/9331-19: Configuration control weaknesses which I adversely affected safety-related plant equipment, were noted in several i
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-9- g ! 1 instances, such as molded case circuit breakers, SDGs, and environmental ! qualification of HOVs.
i 2.3.7 Paragraph 2.3.7, " Functional and Programmatic Weaknesses Could i Adversely Affect the Operability of the Essential Chilled Water System" . ! 2.3.7.1 (0 pen) IFl 498;499/9331-20: Functional and programmatic weaknesses
" ' were observed in the design, testing, modification, and maintenance of the l (essential chilled water system) that, if uncorrected, could adversely affect
< the operability of the system. This and related essential chilled water , system issues are included in Supplement 2 to the CAL. ! l 2.3.7.2 (0 pen) IFI 498;499/9331-21: The ability of the essential chilled ! i water system to function for extended periods, during a design basis accident under low heat load conditions, was never demonstrated, either by testing the i , system at various design basis accident heat loads or by engineering analysis.
. 2 3.8 Paragraph 2.3.8, " Untimely Resolution of Fire Protection issues" ' !
2.3.8.1 (0 pen) IF1 498;499/9331-22: The licensee did not resolve numerous ! ' fire protection issues in a timely manner. The issues included excessive i shrinkage of penetration seals, an unreliable fire alarm system, a large ~ backlog of service requests on fire protection systems, and inadequate control ! of transient combustibles in the plant. i
< 2.4 IFis identified in the Management and Organization Section of the DET j inspection l
2.4.1 Paragraph 2.4.1, " Ineffective Direction and Oversight" . 4 2.4.1.1 (0 pen) IFI 498;499/9331-23: The team concluded that the licensee's- ! - ineffective corrective action processes were major obstacles to improving
plant equipment and human performance. Ineffective ' problem identification, ! ' shallow root cause analyses, inadequate safety evaluat'ons, and lack of .l aggressive problem resolution resulted in short-term rather than long-term
' solutions. i . 2.4.2 Paragraph 2.4.2, " Poor Support and Resource Utilization" , i ' 2.4.2.1 (0 pen) IFI 498;499/9331-24: Staffing levels were marginal or ! 4 insufficient in several key areas. j ~ ' 2.4.3 Paragraph 2.4.3, " Communications And Teamwork Were Weak" , , 2.4.3.1 (0 pen) IFI 498;499/9331-78: Although the Speakout and Employee ! ' Assistance Programs were intended to be anonymous, there was a perception by l . i many employees that they were not.
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, t -10- ! 2.4.3.2 (0 pen) IFI 498;499/9331-25: The threshold of station problem report (SPR) initiation and depth of root cause analyses was not well defined or communicated to staff. ,
2.4.4 Paragraph 2.4.4, " Ineffective Corrective Acticn Precess" l 2.4.4.1 (0 pen) IFI 498;499/9331-26: The team concluded that the licensee's ' ineffective corrective action process was a major obstacle to plant equipment [ and human performance improvement. 2.4.4.2 (0 pen) IFI 498;499/9331-27: Confusion and lack of. training resulted. > in SPRs not being issued in a timely manner on safety-related equipment. 2.4.4.3 (0 pen) IFI 498;499/9331-28: An example of inadequate root cause ' analysis was the licensee's failure to identify the root cause of repeated > failures of SDG fuel injector pump hollow hold-down studs. > 2.4.5 Paragraph 2.4.5, "Inef fective Utilization of Self-assessment and ' Quality Oversight Functior.s" . 2.4.5.1 (0 pen) IFI 498;499/9331-82: Management did not respond effectively ' to the findings, concerns, and recommendations of their principal self- assessment aH quality oversight functions. 2.4.6 Paragraph 2.4.6, " Inadequate Information Systems" No IFIs related to Restart Issues were identified in this paragraph. . ! 3 CAL AND SUPPLEMENT RESTART ISSUES (92701) This section addresses the issues identified in the CAL and its Supplements and assigns issue numbers as listed in Attachment 2. 3.1 Restart Issues Identified in the CAL
3.1.1 The CAL states that the licensee will not restart either unit at STPEGS until the NRC staff has been briefed on the results of the licensee's efforts to correct the overspeed trip condition that was affecting the turbine-driven I auxiliary feedwater pumps. This issue has been identified as Restart Issue-1 ! and is listed in Attachment 2. [ 3.2 Restart Issues Identified in the CAL Supplements { In addition to the issues identified in the CAL, both the CAL Supplements l included additional- topics that the licensee would be required to brief the , NRC staff on prior to restart of either unit at STPEGS. . . 3.2.1 The first bullet of the CAL Supolement (May 7, 1993) pertained to the . ! licensee's SPR process, including process improvements, threshold, and the results of the licensee's review of existing reports for issues affecting j 1
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. 11- ! - operability safe plant operation. This issue has been identified as Restart Issue 2 and is listed in Attachment 2. J 3.2.2 The second bullet of the CAL Supplement (May 7, 1993) pertained to the
SR backlog including reduction accomplished during the current outages and the j licensee's review of outstanding SRs for issues affecting equipment - operability, safe plant operation, and operator _ work-arounds. This issue was ' ' identified as Restart Issue 3 and is listed in Attachment 2. . 3.2.3 The third bullet of the CAL Supplement (Maj 7, 1993) pertained to the > postmaintenance test program, including corrective actions in response to . recent violations and other process improvements and the basis for.the . ! licensee's confidence that equipment removed from service for maintenance, was properly restored to an operable status. This issue was identified as Restart Issue 4 and is listed in Attachment 2. t 3.2.4 The fourth bullet of the CAL Supolement (May 7,1993) included the outstanding design and temporary modifications and other engineering backlog items, including the review of issues affecting equipment operability, safe , plant operation, and operator work-arounds. This issue was identified as
' Restart Issue 5 and is listed in Attachment 2. 3.2.5 The fifth bullet of the CAL Supplement (May 7,1993) addressed the . staffing of the operations department, including the adequacy of current staffing levels, plans for replacing planned and unexpected losses to support safe plant startup and operation, and the adequacy of staffing under emergency
conditions. This issue has been identified as Restart Issue 6 and is listed i in Attachment 2.
' 3.2.6 The sixth bullet of the CAL Supplement (May 7,1993) addressed the status of fire brigade leader training, including verification that this
training meets regulatory requir;ments. This issue has been identified as
Restart Issue 7 and is list ' i Attachment 2. 3.2.7 The seventh bullet f the CAL Supplement (May 7, 1993) addressed the t adequacy of fire protecti, e .puters, including reliability and functionality . of the system. This issue has been identified as Restart Issue 8 and is
listed in Attachment 2. 3.2.8 The eighth bullet of the CAL Supplement (May 7, 1993) addressed the licensee manageinent's effectiveness in identifying, pursuing, and correcting
plant problems, including any plans for independent reviews. This issue has been identified as Restart Issue 9 and is listed in Attachment 2. 3.2.9 The ninth bullet of the CAL Supplement (May 7,1993) addressed _the ' results of the licensee's internal restart readiness reviews. Although this issue was not considered by itself a Restart Issue, related items to this issue, such as the line management assessment plan, have been separately identified as IFIs and will be reviewed as part of the restart inspection process. . I r e
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! 3.2-.10 The-first bullet of the second CAL Supplement (October 15,1993) ! pertained to the effectiveness of- the licensee's SPEAK 0UT program in addressing employee safety concerns. This issue has been identified as
Restart Issue 10 and is listed in Attachment 2. , 3.2.11 The second bullet of the second CAL Supplement (October 15, 1993) pertained to topics associated with standby diesel generator reliability. . . This issue has been identified as Restart Issue-11 and is listed in Attachment 2. 3.2.12 The third bullet of the second CAL Supplement (October 15, 1993) pertained to essential chiller reliability issues. This issue has been identified as Restart issue 12 and is listed in Attachment 2. 3.2.13 The fourth bullet of the second CAL Supplement (October 15, 1993) addressed the licensee's system certification program. This issue has been identified as Restart Issue 13 and is listed in Attachment 2. 3.2.14 The fifth bullet of the second CAL Supplement (October 15,1993) pertained to the reliability and operability of the feedwater_ isolation bypass valves associated with both units. This issue has been identified as Restart issue 14 and is listed in Attachment 2. 3.2.15 The sixth bullet of the second CAL Supplement (October 15, 1993) addressed the adequacy of tornado damper testing. This issue has been identified as Restart issue 15 and is listed in Attachment 2. 3.2.16 The seventh bullet of the second CAL Supplement (October 15, 1993) addressed the effectiveness of emergency preparedness personnel accountability. This issue has been identified as Restart Issue 16 and is listed in Attachment 2. 4 IFIs IDENTIFIED IN THE ORP (92701) This section was structured to address items related to Restart Issues identified in the licensee's ORP. Several of these IFIs are similar in character and scope to previously identified items related to Restart Issues, and the matrix of common items in Attachment 3 was included to identify and cross-reference common items. 4.1 (Open) IFI 498;499/9331-29: A new methodology will be developed to properly characterize the existing maintenance backlog and newly generated- SRs. 4.2 (0 pen) IFI 498:499/9331-30: Additional backlog reduction goals for resumption of power operation established for engineering evaluations are:
- . .- . . -13- Demonstrate progress on completing a general backlog reduction 'from a + peak value of approximately 1400 items down to 600 items by the end of 1993. No Operating Experience Reports, SPRS, Design Change Requests, Document
Change Notices, or nondesign change Plant Change Forms (PCFs) greater than 1 year old without an engineering evaluation. , 4.3 (0 pen) IFI 498;499/9331-31: Additional backlog reduction goals established for administrative / programmatic changes are: Update the 311 vendor documents with five or more open amendments that
have been identified by Operations and Maintenance as impacts on their performance. Update key control room design drawings.
No Master Parts List Change Forms open greater than 60 days.
Demonstrate progress on reducing PM and SR history backlogs from 6100 to
200 by the end of 1993. 4.4 (0 pen) IFI 498;499/9331-32: The effectiveness of the Department Management Team and the Site Management Team will be periodically evaluated by STPEGS senior management. 4.5 (0 pen) IFI 498;499/9331-33: Technical Services will further support Operations by qualifying more personnel as fire brigade leaders. 4.6 (0 pen) IFI 498;499/9331-34: Power ascension will be coordinated by Power Ascension Test sponsors reporting directly to the plant manager. 4.7 (0 pen) IFI 498;499/9331-35: The certification and acceptance process will be procedurally controlled and documented by two procedures. One procedure will define a comprehensive package that demonstrates each key system has been adequately reviewed and any outstanding items have been appropriately evaluated and dispositioned. A second procedure will be developed that will require a comprehensive walkdown followed by acceptance of the system by the plant manager. 4.8 (0 pen) IFl 498;499/9331-36: Senior shift managers will provide continuous management representation and presence during selected evolutions throughout the power ascension program. The senior shift manager's primary function will be to ensure that the exercise of command and control authority of licensed operators is not diluted by the increased level of activities. The senior shift manager will also be responsible for assessing the conduct of Operations, Maintenance, and other support groups.
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. > -14- ' i ? 4.9 (0 pen) IFI 498;499/9331-37: The goal for Unit I and common power block l SRs is below 1000. . 4.10 (0 pen) IFl 498;499/9331-38: The goal for Unit 2 power block SRs is ! below 850. 4.11 (0 pen) IFl 498;499/9331-39: There will be no outstanding SRs that- ' adversely affect plant safety or reliability (Priority 1 and 2). 4.12 (0 pen) IFI 498;499/9331-40: All Engineering backlog items that do not meet these general criteria will be completed:
Open engineering work items that do not adversely. affect plant material I
condition as determined by the size and age of the carticular backlog. l ! Open engineering work items that have no significant impact on safety- l + ' related equipment or system operability. i 4.13 (0 pen) IFI 498;499/9331-41: Additional backlog reduction goals for
resumption of power operations established for design / physical changes were:
Reduction in the number of undispositioned s. <nforming PCFs to less
than 50 that are greater than 30 days old. , Reduction to 15 Temporary Hodifications from the current level of l
24 installed for greater than 6 months for Unit I and Common. 4.14 (0 pen) IFI 498;499/9331-42: The additional backlog reduction goal ! established for carryover items from past programs was to either complete the l engineering work product or convert the item to a current work program. , 4.15 (0 pen) IFI 498;499/9331-43: The turbine-driven auxiliary feedwater i pumps will be subjected to an augmented surveillance program that will confirm the reliability of the equipment. 4.16 (0 pen) IFl 498;499/9331-44: Engineering will perform calculations , related to the essential chilled water system, which will provide the basis for evaluation and analysis of minimum and maximum chilled water loads under a range of weather-related conditions and postulated design basis accidents.
Strategies will be developed to operate and test the system, and an evaluation i of proposed chiller enhancements will be completed. ! i ' 4.17 (0 pen) IFI 498;499/9331-45: Ensure that the essential chillers were
capable of performing their design function. ! , 4.18 (0 pen) IFI 498;499/9331-46: The status of the Technical Support Center ! ' diesel will be evaluated as part of the assessment process prior to the ' resumption of power operation. ! ! 1 i i , ,. . , . . - - ,, _ , . , . . . . - - - - - , -
, - -. ..- . u. . __ .. ( ! - . -15- f 4.19 (0 pen) IFI 498;499/9331-47: All SRs involving automatic functions will be evaluated and necessary work performed to ensure that no significant impact on system operability or operator burden exists. Any remaining inoperable automatic functions will be analyzed in the aggregate to ensure safe and- reliable plant operation will not be unacceptably impacted. 4.20 (0 pen) IFI 498;499/9331-48: .The status of the solenoid-operated valve , issues will be evaluated as part of the assessment process. . 4.21 (0 pen) IFI 498;499/9331-49: Management will review the number of components on increased surveillance testing frequency to ensure that the burden on operations and maintenance relating to the testing of these components will not adversely affect the safe operation of the plant. 4.22 (0 pen) IFI 498:499/9331-50: The plant modification for permanent flow instrumentation in the auxiliary feedwater system will be installed. 4.23 (0 pen) IFl 498;499/9331-51: Precision calibrations were being performed on the instal',ed flow instruments in the component cooling water heat exchanger out.et to the essential cooling water system to preclude having to use temporary flow instruments. , 4.24 (0 pen) IFI 498;499/9331-52: Design Changes or PCFs have been initiated to eliminate the use of temporary flow instruments or temporary pressure gages in the surveillance tests for the following systems: Essential Chilled Water, High Head Safety Injection, Spent Fuel Pool Cooling, and Screen Wash Booster - Pumps. The status of these changes will be evaluated as part of the assessment. , 4.25 (0 pen) IFI 498;499/9331-53: The assessment process will evaluate and determine the acceptability of continued operation at specific milestones including: prior to Mode 4; prior to criticality; power ascension above 50 percent power; completion of the first week of full power operation; after maintaining full power for 1 month; as determined thereafter. The process will include involvement of line and senior management, station assessment oversight groups (e.g., Nuclear Assurance, Independent Safety Engineering ' Group, Plant Operations Review Committee (PORC), Nuclear Safety Review Board (NSRB), Planning & Assessment) the Operational Readiness Review Panel (ORRP), outside consultants, and industry groups as determined by STPEGS executive management. . I 4 4.26 (0 pen) IFI 498;499/9331-54: In order to assure a consistent and
integrated approach to the internal assessmcat process, a Line Management j Assessment Plan will be prepared by line management and approved by the Group , Vice President, Nuclear prior to core reload. Conceptually, the plan will ! consist of the following elements: , Line managers with assessment responsibilities ass)ciated with
resumption of operations will be defined. f e . _ - . .
- ..- - - l ' ! - -16- , ! < Assessment points or plateaus will be defined from core reload to i
100 percent reactor power.
Line managers designated above will prepare self-assessment
' checklists / plans for their functional area for each of the assessment milestones. These checklist / plans will address the hardware, program, , and performance issues described in this ORP, including issues described , - in the NRC's CAL and CAL Supplement and appropriate issues for the , diagnostic evaluation response. The checklists will both review performance to date and readiness to proceed to the next plateau. Prior to advancing beyond any assessment milestone, the checklists for
that plateau will be completed and reviewed. The PORC will review the , checklists for those functions that directly report to or support the Plant Managers organization. The ORRP will review the results of the PORC review and checklists from functions that are outside the plant manager's organization; e.g., Nuclear Licensing. > 4.27 (0 pen) IFI 498;499/9331-55: In addition to the Line Management Assessment Plan, an Independent Assessment Plan will also be prepared and approved prior to core reload. Conceptually, this plan will address the following: An integrated surveillance / observation plan for internal (e.g., Quality '
Assurance [QA], Assessments, and Independent Safety Engineering Group) and external industry groups for specific plant events. This process will be managed and coordinated by the Nuclear Assurance Department. A review of the Line Management Assessment Plan at each assessment
pl ateau. The plan will contain specific criteria for the assessment of the process. The General Manager-Nuclear Assurance will give the Group Vice President, Nuclear an independent opinion on readiness to proceed , to the next plateau as an input on each plateau decision. i Prior to ascending to Mode 4 and prior to criticality, the independent
members of the NSRB will conduct a review of the decision / assessment . process to date and provide that input to the General Manager-Nuclear
Assurance. After reaching 100 percent power, the General Manager-Nuclear Assurance '
will conduct a critique of the independent assessment process and provide a report on the lessons learned and recommendations on ~; improvements for the follow-on unit. Collectively, the Plant Manager and ORRP will report to the Vice
President, Nuclear Generation and the Group Vice President, Nuclear on i performance to date and readiness to proceed to the next plateau. They will also provide a recommendation on whether or not to proceed to the , next plateau. , , , - - _ , y
- --
. I i . ! -17- , The Vice President, Nuclear Generation will recommend and the Group Vice
President, Nuclear will approve proceeding to the next plateau. , After reaching 100 percent power, an overall critique of the process
will be conducted by the Vice President, Nuclear Generation. Lessons learned and recommendations from that critique will be factored into the
! plan for the follow-on unit. , ! 4.28 (0 pen) IFI 498;499/9331-56: Major changes were underway within the operational organizations in Nuclear Generation. The thrust was to unitize the Operations and Maintenance organizations to provide more organizational - focus and to shorten communication chains within the organization. 4.29 (0 pen) IFI 498;499/9331-57: A six-crew operating schedule will be i implemented. Each crew will consist of five licensed and five nonlicensed operators. ' 4.30 (0 pen) IFI 498;499/9331-58: A program modification to the fire protection computer revising over 1000 alarm messages to a more user friendly - format which provides the type and location of each alarm device and i ' automatically provides a hard copy printout of the associated Fire Pre-Plan document number for ease of reference will be installed. ! 5 4.31 (0 pen) IFI 498;499/933159: To enhance operational emphasis on safety- related and power block operations and to further reduce the burden on the ' operators, the responsibility for nonsafety-related support systems outside l f the protected area will be transferred to the Technical Services Department, 4.32 (0 pen).IFI 498;499/9331-60: Operators will receive specific training ! which, as a minimum, will consist of: Performance on the simulator of a reactor and plant startup from Mode 4
to turbine roll with performances of surveillances and malfunctions. l At-power operation casualties will be included. i ! Training on modifications made during the outage. i
4.33 (0 pen) IFI 498;499/9331-61: Typically assign two supervisors for each maintenance crew: one supervisor to provide enhanced field management of the crew and the other to plan the work for the next week. ! 4.34 (0 pen) IFI 498;499/9331-62: Criteria for Maintenance Effectiveness and Material Condition i No outstanding SRs that affect unit safety or reliability - No l
Priority is or 2s.
l 4 . _ _
. . . - - __ _ , _ , 9
' . b -18- , i. Demonstrate ability to manage maintenance workload - Total open SRs !
meets goal (less than 1000 in Unit 1) and workoff rate trend remains i positive. , Changes in SR generation rate are evaluated and understood to ensure
threshold for deficiency identification was acceptable - (SR generation rate is consistent with plant condition). , PM deferrals analyzed and corrective actions in progress - Goal (less
than 20) met and trend remains positive. t Main Control Board deficiencies - Goal (less than 10) met and trend.
remains positive. Inoperable automatic control functions - Aggregate does not adversely
affect operations ability to perform quality rounds and handle normal work load. Positive trend continuing in resolving inoperable functions. ! ' 4.35 (0 pen) IFI 498;499/9331-63: A test system for the system performance software will be initialized onsite. 4.36 (0 pen) IFI 498;499/9331-64: The ,verall ability of Nuclear Engineering
management to manage the required work load has beer, enhanced through assignment of a new Vice President, Nuclear Ergineering, with further , improvement following a comprehensive realignment of the Engineering
organization to be completed prior to resumption of power operation. 4.37 (0 pen) IFI 498;499/9331-65: During the plant startup, Engineering will
provide 24-hour on-shift support to facilitate effective interface between 2 Operations, Maintenance, and Engineering. This will be accomplished by '
24-hour staffing of the Technical Support Engineering organization. The on- , shift staff will have direct access to Design Engineering and other startup , components of the Engineering organization on an as-needed basis throughout
the startup phase. i . 4.38 (0 pen) IFI 498;499/9331-66: As part of the effort to reduce the burden ! on the Operations staff and to allow them to focus on the power block, the Technical Se / ices Department will be assuming the responsibility for certain '
tasks that a.e outside the protected area. 4.39 (0 pen) IFI 498;499/9331-67: STPEGS will ensure that line management ownership of the corrective action process is established, necessary
enhancements to the SPR process are implemented and proven, and any existing
backlog of SPR actions were assessed for potential impact on equipment operability and safe plant operation.
4.40 (0 pen) IFI 498;499/9331-68: The " Post-Maintenance Test" program was ' restructured to consolidate program information and to better define and communicate testing requirements, j ! ! r 7tr -g y
. .. ( . - -19- l . 5 NRC SPECIAL AND R0llTINE INSPECTION REPORTS AND NRC STAFF ACTIONS (92701) 5.1 Previously Identified Unresolved Items. IFis. and Violations in Routine and Special Inspection Reports
' 5.1.1 (0 pen) IFI 498;499/9116-02: Operator overtime issues. '
5.1.2 (0 pen) IFI 498;499/9214-03: SDG availability issues. 5.1.3 (0 pen) IFI 498;499/9221-03: SDG availability issues. - 5.1.4 (0 pen) IFI 498;499/9224-03: Essential Chiller reliability and
unavailability issues. , 5.1.5 (0 pen) IFI 498;499/9208-01: Reactor coolant system overcooling. 5 5.1.6 (0 pen) Violation 499/9226-03: Failure to perform an adequate postmaintenance test. 5.1.7 (0 pen) Violation 498;499/9235-02: Failure to initiate an SPR. 5.1.8 (0 pen) Violation 499/9304-03: Failure to maintain minimum control room shift staffing. 5.1.9 (0 pen) URI 499/9315-03: Cause of high fuel oil strainer differential pressure unknown. 5.1.10 (0 pen) IFI 498;499/9324-01: Feedwater check valve and isolation bypass valve leakage.
5.1.11 (0 pen) URIs 498;499/9319-01 through -07: Feedwater isolation bypass , valve issues. 5.1.12 (0 pen) IFI 498;499/9306-07: This IFI concerned the opening , differential pressure trace of AISIMOV0001B, the pressurizer power-operated , relief valve associated with both units. _; , 5.1.13 (0 pen) Violation 498;499/9217-02: This violation concerned the failure of cognizant licensee personnel to inform the control rooms that both l units were in Technical Specification 3.0.3. , i 5.1.14 (0 pen) Violation 498;499/9217-04: Failure to follow procedures in the ! issuance of guidance pertaining to the Technical Specifications. 5.1.15 (0 pen) Violation 498;499/9224-01: Failure to take adequate corrective action to preclude essential chill water switch malfunctions during valving-in processes following maintenance. 5.1.16 (0 pen) Violation 498;499/9235-06: Two examples of fire protection i violations. j 1 l ! - - - - - - - - - .
- '" . . . _ . _ _ . . . .. 4 -20- 5.1.17 (0 pen) Violation 498;499/9303-01: Eight examples of station personnel self-verification problems. 5.1.18 (0 pen) IFI 498;499/9304-04: Reactivity management followup item concerning operation of the boron thermal regeneration system. 5.1.19 (0 pen) Violations 498;499/9305-01, 04, 05, 07: These violations ' concern the inoperability of the turbine-driven auxiliary feedwater pump and the emergency diesel generators. ,
5.1.20 (0 pen) Violation 499/9308-02: Violation of the Technical- Specifications for Valve SI-31A being inoperable for an entire fuel cycle. 5.1.21 (0 pen) Violation 498/9308-04: Inadequate corrective action performed t resulting in numerous motor-operated valve problems. 5.1.22 (0 pen) Violation 498;499/9309-01: Transient combustibles not stored in accordance with licensee requirements. 5.1.23 (0 pen) Violation 498/9311-04: Reactivity management violation resulting from a failure to maintain an operable boron injection flow path as required by the Technical Specifications. 5.1.24 (0 pen) Violation 499/9315-01: Technical Specification violation concerning a residual heat removal pump that was restored to service with the t incorrect reference data being used in previous surveillance testing.
5.1.25 (0 pen) Violation 498/9320-02: Technical Specification violation
concerning solid state protection system testing being conducted without the t latest procedural revisions included in the surveillance package. l 5.1.26 (0 pen) Violation 498/9321-01: Corrective action violation concerning , failures to take prompt action following the discovery that seismic fasteners were missing on the card cages and power supply racks of the Qualified Display Parameter System.
5.1.27 (0 pen) Violation 499/9322-01: Two examples of operations personnel l failing to follow procedures resulting in a loss of spent fuel pool cooling - for over 13-hours. 1 5.1.28 (0 pen) Violation 499/9322-02: Corrective action violation concerning the failure to take action to preclude recurrence of safety-related valves changing position during break-before-make bus transfers. 5.1.29 (0 pen) URI 498;499/9325-02: An item unresolved pending further NRC review concerning station accountability during day shifts. 5.1.30 (0 pen) LER 498/92-04: Concerns the failure to adequately test the shunt trip coil of the reactor trip breakers.
1 '
.- . -- . _ _ . - ! -21- , ! i 5.1.31 (0 pen) LER 498/92-07: Unplanned engineered safety features actuation , due to inadequate surveillance test. 5.1.32 (0 pen) LER 498/92-14: Unplanned engineered safety features actuation due to inadequate surveillance test. 5.1.33 (0 pen) LER 498/92-16: Unplanned engineered safety features actuation ! ' due to inadequate surveillance test. 5.1.34 (0 pen) LER 498/92-20: Toxic gas monitor not in correct condition for
plant operations due to operator error.
? 5.1.35 (0 pen) LER 498/93-05: Emergency Diesel Generator 13 failed to start - on demand due to paint fouling the fuel metering rods. l i 5.1.36 (0 pen) LER 498/93-07: LTurbine-driven auxiliary feedwater pump inoperable due to repetitive overspeed trips. 5.1.37 (0 pen) LER 498/93-17: Feedwater isolation bypass valve inoperable due to positioner and solenoid equipment being beyond its qualification life. j t' 5.1.38 (0 pen) LER 498/93-20: Feedwater isolation bypass valve inoperable due to inadequate closing pressure forced the valves to open at normal operating
pressures. 5.1.39 (0 pen) LER 499/92-04: Technical Specification 3.0.3 entry due to Target Rock solenoid operated-containment isolation valves failing to close. l 5.1.40 (0 pen) LER 499/93-04: Reactor trip due to the startup steam generator feedwater pump failing to operate because of recurrent problems with water
intrusion into the pump's lube oil system. , 5.1.41 (0 pen) LER 499/93-05: Control room unmanned by a senior reactor i operator during Mode 4 operations. 5.1.42 (0 pen) LER 499/93-12: Loss of spent fuel pool cooling for 4 approximately 13-hours. , 5.2 IFis Related to Office of Nuclear Reactor Regulation (NRR) Actions 5.2.1 (0 pen) IFI 498;499/9331-69: The licensee's response to Bulletin 88-08, , " Thermal Stresses in Piping Connected to Reactor Coolant Systems." 5.2.2 (0 pen) IFl 498;499/9331-70: This IFI concerned the licensee's , commitment to revise the Technical Specifications that require specific levels ) of boron concentration in shutdown margin calculations. , 5.2.3 (0 pen) IFl 498;499/9331-71: Concerns the licensee's commitment to f revise the Technical Specifications concerning the surveillance requirements , of turbine-driven auxiliary feedwater pump testing. J
_ _ . . . _ . _ . ._ - , ' . . 4 -22- 5.2.4 (0 pen) IFl 498;499/9331-72: Concerns additional information requested by NRR in the licensee's initial response to Generic letter 93-04.. , 5.3 IFIs identified from NRC Staff Actions 5.3.1 (0 pen) IFI 498;499/9331-73: NRC will assess the operating staff workload issues at STPEGS and licensee management's action to resolve these ~ staffing issues. 5.3.2 (0 pen) IFI 498;499/9331-74: NRC will assess the licensee's engineering analysis for essential chiller operation under low heat load accident conditions. . 5.3.3 (0 pen) IFI 498;499/9331-75: NRC will assess the licensee's action to' resolve fire protection deficiencies at STPEGS. These deficiencies include: (1) fire protection computer alarm system and operator training on the system;
(2) a large backlog of SRs on fire protection systems; (3) control of ' transient combustibles in the plant; and (4) Fire Brigade leader . qualifications and the impact on operations staffing. { ! 5.3.4 (0 pen) IFI 498;499/9331-76: Failure of tornado dampers could prevent cooling of safety-related components and systems. Thirty dampers had not been tested to verify their designed operation. The. licensee agreed to test the dampers. NRC will evaluate the licensee's test procedures and results. . .
, t t i i ! ! i i ! ! ! )
? _.-- _ . _ , . . .
= - - _. __ . ~ . ' . ' ATTACHMENT 1
-, 1 PERSONS CONTACTED , 1.1 Licensee Personnel , J. Sheppard, General Manager, Nuclear Licensing ~ M. Coughlin, Senior licensing Engineer , other members of the licensee's staff ' 1.2 NRC Personnel W. Johnson, Chief, Project Section A, Division of Reactor Projects M. Satorius, Project Engineer, Project Section A, Division of Reactor Projects T. McKernon, Reactor Inspector, Operations Section, Division.of Reactor Safety 2 EXIT MEETING A telephonic exit meeting was conducted on October 8, 1993. During this ' meeting, the inspectors reviewed the scope and findings of this report. The licensee did not identify as proprietary any information provided to, or -; reviewed by, the inspectors. , !
s O C t ~ ! , , - - - - - -
. . , ATTACHMENT 2 RESTART ISSUES /RELATED ITEMS MATRIX ' RESTARTISSUE RELATED ITEMS 1 Turbine-driven Auxiliary Feedwater Pump Reliability 9331-07, 08, 09, 10, 43, 50, 71 and Testing Methodology 9305-04, 05, 07 Unit 1 LER 9307 Unit 2 LER 9304 2 Station Problem Report Process, Threshold, 9331-06, 18, 23, 25, 26, 27, 28, 67 Licensee's Review of Existing Reports for Issues 9235-02 , Affecting Operability and Safe Plant Operation 9224-01 9321-01 9322-02 9308-02, 04 3 Service Request Backlog, Including Reduction 9331-02, 03, 07, 08, 09, 29, 31, 37, 38, 39, Accomplished During the Current Outages and the 47, 49, 62, 79, 80 Licensee's Review of Outstanding SRs for Issues Affecting Equipment Operability, Safe Plant Operation, and Operator Work-arounds 4 The Postmaintenance Test Program, Including 9331-03, 04, 07, 10, 13, 14, 15, 51, 63, 68, Corrective Actions in Response to Violations and 79 Other Process improvements and the Basis For the 9226-03 Licensee's Confidence That Equipment Removed From 9320-02 Service for Maintenance is Properly Restored to an 9305-01, 05, 07 , Unit 1 LER 9204, 9207, 9214, 9216, 9305 Operable Status 1 of 3 . t , , . , , , - ,y -m-e ., -.--,,,-,w,. , , - - - ar, -,, c + - - +<n, , _ .,n,.- ,.,___ --
. > .. . RESTARTISSUE RELATED ITEMS 5 The Outstanding Design Modifications, Temporary 9331-02, 04, 08, 12, 16, 18, 19, 20, 21, 30, Modifications, and Other Engineering Backlog Items, 31, 40, 41, 42, 44, 45, 48, 52, 64, 65, 77,81 Including the Licensee's Review of These For Issues 9208-01 Affecting Equipment Operability, Safe Plant 9306-07 Operation, and Operator Work-arounds 9315-01 Unit 1 LER 9220 Unit 2 LER 9204 6 Adequacy of Operations Staffing 9331-01, 03, 24, 56, 57, 59, 60, 65, 66, 73 l 9116-02 9304-03, 04 9311-04 9322-01 Unit 2 LER 9305 Unit 2 LER 9312 7 Adequacy of Fire Brigade Leader Training and 9331-04, 33, 75 Qualifications 8 Adequacy of Fire Protection Computers and Software, 9331-02, 04, 17, 22, 58, 75 the Licensee's Success in Reducing the Number of 9235-06 Spurious Fire Protection System Alarms, and Other 9309-01 Fire Protection Hardware Problems 9 Licensee Management's Effectiveness in Identifying, 9331-04, 05, 06, 17, 18, 22, 23, 25, 32, 34, Pursuing, and Correcting Plant Problems 35, 37, 46, 54, 55, 56, 61, 62, 65, 67, 69, 70, 72, 73, 80, 82 9321-01 9322-02 9224-01 9217-02, 04 9303-01 9308-02, 04 Unit 1 LER 9204 2 of 3 . t I . . . . - - . . . __, - .- _ ., - , , ,.-e _. - . . _. ,s ,
,
, RESTART ISSUE RELATED ITEMS 10 NRC Review of the Effectiveness of the Licensee's 9331-78 SPEAK 0UT Program 11 Standby Diesel Generator Reliability 9331-08, 09, 11, 12, 13, 16, 19, 28 9214-03 9221-03 9305-01 9315-03 Unit 1 LER 9305 12 Essential Chiller Reliability 9331-10, 13, 20, 21, 44, 45, 74 9224-03 13 Monitoring of the licensee's System Certification 9331-35, 53 Program 14 Adequacy of the Licensee's Resolution of the 9319-01 through 07 Reliability and Operability of the Feedwater 9324-01 Isolation Bypass Valves Unit 1 LER 9317 Unit 1 LER 9320 15 Tornado Damper Issues 9331-76 16 Emergency Preparedness Accountability Issues URI 498;499/9325-02 3 of 3 . f . . . _ . - _, .~ . - - , _ _ . , . . . . . . . _,
_ _ . . , 't ATTACHMENT 3 < SUMMARY OF INSPECTION FINDINGS AND COMMON ITEMS IFI/URl/VIOLAT!ONS REFERENCE IFI Number Descriptor Report Section Common Items (IFI,URI,VI0s.etc) 9331-01 Adverse impact to control 2.1.1.1 9331-36, 57, 59, 65, 66, 73 room staff by workload and site support 9331-02 Operators affected by 2.1.1.2 9331-06, 29, 35 degraded equip.; equip. . walkarounds, etc. 9331-03 Focus and overview of Plt. 2.1.1.3 9331-36, 49, 53, 56, 57, 65, 73 operations not effectively maintained by SS/CR staff 9331-04 Poor support to Operations 2.1.2.1 9331-05, 24, 36, 49, 53, 54, 55, 57, 65, 73 ' 9331-05 Confusing and conflicting 2.1.3.1 9331-25, 27, 36, 56 mgt. guidance to Ctrl. Rm. staff 9331-06 Ineffective Ngt. support to 2.1.5.1 9331-18, 23, 26, 37, 38, 39, 40, correct program & component 41, 42, 45, 46, 47, 48, 50, 51, problems 58, 62, 75 9214-03, 9221-03, 9224-03, 9315-03, 9324-01, 07; 9306-07 1 of 8 , ! ! . .. . .- - . .- . . . . . . . . . . . . . . - . . . . . . . . .
,, . IFI Number Descriptor Report Section Common Items (IFI,URI,VI0s,etc) 9331-07 M & T weaknesses reduced 2.2.1.1 9331-10, 13, 14, 20, 35, 43, 44, reliability of pit. equip. 68, 71 ' 9235-02 9226-03 9331-08 Ineffective cms & PMs 2.2.1.2 9331-9, 16, 17, 20, 22, 37, 38, contributed to poor equip. 39, 47, 62, 75 performance 9324-01, 07 9331-09 Ineffective cms resulting 2.2.1.3 9331-11, 12, 17, 18, 23, 25, 28, I from inadequate RCAs, work 64, 75 control & craft performance 9331-10 Equip. operability not 2.2.1.4 9331-07, 13, 14, 15, 20, 43, 63, always verified by sury. & 68, 71 PMTs 9226-03 L .. 9331-11 Poor RCA for SDG hold down 2.2.1.5 9331-09, 13, 18, 19, 28; 9214-03; bolts 9221-03 9331-12 Configuration control / 2.2.6.1 9331-08, 09, 16, 18, 19, 25, 28, engineering communication 64 for SDG hold down bolts 9331-13 PMT program and 2.2.7.1 9331-07, 10, 14, 20, 68 implementation weaknesses 9226-03 9320-02 9303-01 4 9305-01, 05, 07 Unit 1 LER 9207, 9214, 9216, 9305 9331-14 Inadequate PMT manual 2.2.7.2 9331-07, 13. 20, 58; 9226-03 9331-15 TS Sury. program and 2.2.8.1 9331-10, 49, 71 procedures need enhancement " 2 of 8 . 4
. 4 - .,. . _ , . . . - . - . , _ , . - - . . . , . . . _ _ , . . - . - -~ - <-- , . _ , - - .. -.m . ., . . , , _ - - . , ,-c . m._,., .-., _ ... - - . . , -
.. - . .. . IFI Number Descriptor Report Section Common Items (IFI,URI,VI0s,etc) 9331-16 Configuration control 2.3.1.1 9331-12, 18, 19, 20, 21, 28, 34 , weaknesses affected safety- related equipment 9331-17 Fire protection issues not 2.3.1.2 9331-02, 04, 22, 58, 75 resolved in a timely manner 9235-01 9309-01 9331-18 Engrg. RCA & corrective 2.3.1.3 9331-11, 12, 16, 21, 23, 26, 28, actions weak 33, 40 9331-19 Configuration control 2.3.6.1 9331-12, 16, 20, 22, 41, 44 weaknesses 9331-20 Weaknesses in essen. CHW 2.3.7.1 9331-22, 44, 45, 46, 52, 74 system design, testing, 9224-03 ' mods. & maintenance 9331-21 Design analysis of the 2.3.7.2 9331-20, 44, 45, 46, 74; 9224-03 1 essen. CHW sys. under DBA not performed 9331-22 Fire protection issues not 2.3.8.1 9331-17, 59, 76; 9235-01; 9309-01 resolved 9331-23 Ineffective corrective 2.4.1.1 9331-09, 11, 17, 18, 26, 27 action processes 9331-24 Staffing levels marginal or 2.4.2.1 9331-01, 02, 03, 24, 49, 56, 57, insufficient 59, 60, 65, 66, 73 9116-02; 9208-01; 9304-03, 04; 9311-04; 9322-01 Unit 2 LERs 9305, 9312 3 of 8 . 6 9 _ . _ _ _ _ _ . _ . . _ _ . _ . _ __ . . . _ _ . - - _ . . _ . , _ . . _ _ . _ _ _ _ _ - . _ . - _ . - _ _ , , . _ , . . _ . . . _ . . . . . . _ , . _ . _ _ , _ . . ... . . - _ _
. _ . . . , IFI Number Descriptor Report Section Common Items (IFI,URI,VI0s,etc) 9331-25 Threshold of SPRs 2.4.3.7 9331-06, 18, 23, 27, 67 9235-02; 9224-01; 9321-01 ' " 9322-02 EA 93-047; EA 93-057 9331-26 Ineffective corrective 2.4.4.1 9331-06, 08, 11, 18, 22, 23, 25,
' action process 28, 61, 62, 67 t 9331-27 SPR training & confusion 2.4.4.2 9331-06, 23, 25, 67 9235-02 9331-28 Inadequate RCA for SDG fuel 2.4.4.3 9331-11, 12, 16, 18, 19, 23, 25, injector hold down bolts 30, 67 9331-29 Develop new method to 4.1 characterize maintenance backlog 9331-30 Engineering backlogs 4.2 reduced 9331-31 Additional backlog 4.3 reduction goals 9331-32 Effectiveness of dpt. mgt. 4.4 team periodically evaluated 9331-33 Tech. Services will qualify 4.5 more fire brigade leaders 9331-34 Power ascension coordinated 4.6 9331-35, 36, 49, 53, 54, 55, 65 , 9331-35 System Certification and 4.7 9331-36, 37, 38, 43,.46, 47, 49, Acceptance Process 53, 62, 71, 75 9214-03; 9222-03;-9224-03; 9319-01 . . 4 of 8 . . - . . -. . . .- . . . . - - . . - _ _ - - - _ . _ . - - . _ , _ . _ - . . . . . . - ~ . - . _ , _ - - - - - - . -- -
- . . - ., i ! IFI Number Descriptor Report Section- Common Items (IFI,URI,VIDs,etc) 9331-36 Senior Shift Managers 4.8 9331-01, 03, 04, 24, 53, 54, 55, provide continuous mgt. 56, 59, 73 repr. and presence 9331-37 SRs below 1000 Unit 1 and 4.9 9331-31, 38, 39, 47, 75 common power block 9331-38 SRs below 850 Unit 2 power 4.10 9331-31, 39, 47, 75 block 9331-39 No outstanding Priority 1 & 4.11 9331-31, 37, 38, 47, 75 . ' 2 SRs 9331-40 Engineering backlog items 4.12 9331-33, 41, 42, 45, 46, 47, 50, 51, 52, 58, 74, 76 9331-41 Additional backlog 4.13 9331-31, 42, 45, 46, 47, 50, 51, reduction items 52, 58, 74, 76 9331-42 Carryover items from past 4.14 9331-31, 41, 45, 46, 47, 50, 51, . programs 52, 58, 74, 76 9331-43 TDAFW augmented 4.15 9331-07, 13, 15, 35, 49, 50, 71 surveillance program 9305-04. 05 9331-44 Engineering cales. for the 4.16 9331-16, 20, 21, 40, 45 essen. CHW system, tests and proposed enhancements 9331-45 Ensure essen. chillers 4.17 9331-10,-13,.20, 21, 40, 44, 21, perform their design 74 function 9224-03 9331-46 Status of the TSC diesel 4.18 9331-33, 40, 35 evaluated 1 i 5 of 8 ' t ' r - . . . . . . . . x .-- ~ . . - . . . . . . . . .
. . , IFI Number Descriptor Report Section Common Items (IFI,URI,VIDs,etc) 9331-47 SRs involving automatic 4.19 9331-31, 37, 38, 39, 35 functions 9331-48 SOV issues evaluated 4.20 9331-30, 35, 40, 62 9331-49 Mgt. review of components 4.21 9331-15, 35, 43, 71 on increased number of surveillances . 9331-50 Install of AFW permanent 4.22 9331-30, 40, 52 ' flow instrumentation 9331-51 Calibrations of the CCW 4.23 9331-06, 18, 30, 40 heat exchangers flow > instruments 9331-52 Evaluation of design 4.24 9331-30, 40, 50 changes to eliminate temporary flow instruments 9331-53 Continued operation 4.25 9331-03, 04, 55, 35, 36 assessed at different milestones 9331-54 Line management Assessment 4.26 9331-03, 04, 35, 53, 55, 56 Plan 9331-55 Independent Assessment Plan 4.27 9331-03, 04, 53, 54, 56 9331-56 Changes to improve 4.28 9331-01, 02, 03, 05, 13, 15, 25, communications 27, 36, 49, 61, 65, 67, 73, 79 9331-57 Six-crew operating schetule 4.29 9331-01, 03, 24, 73 implemented 9331-58 Fire protection computer 4.30 9331-75, 37, 38, 22, 17, 08 modification 6 of 8 . , - , . w , . ,, - ..r. . . , . . , , . , , . - . . e
. . , . IFI Number Descriptor Report Section Common Items (IFI,URI,VIDs,etc) 9331-59 Non-safety-related support 4.31 9331-64, 66 systems transferred to Tech. Services Opt. 9331-60 Operator Training 4.32 9331-56, 27, 25, 04 9331-61 Two supervisors for each 4.33 9331-62, 67, 56, 13, 09, 08, 06 mnt. crew 9331-62 Maintenance effectiveness 4.34 9331-61, 67, 68, 37, 38, 39, 25, and mtl. condition criteria 26, 13, 10, 09, 08, 07 9331-63 Test system for system 4.35 9331-56, 26, 23, 18, 10, 06 performance software 9331-64 Realignment of engineering 4.36 9331-56, 40, 24, 06 organization 9331-65 24 Hour on-shift support to 4.37 9331-73, 56, 36, 34, 26, 23, 18, operations by engineering 06, 04 9331-66 Tech Services Dpt. assuming 4.38 9331-65, 64, 40, 06, 04 more responsibilities 9331-67 Line mgt. ownership of the 4.39 9331-62, 61, 56, 49, 27, 26, 25, corrective action process 23, 10, 09, 08, 07 9331-68 PMT program restructured 4.40 9331-62, 14, 13. 10, 07 9331-69 Bulletin 88-08 Response 5.2.1 9331-30, 40 " Thermal Stress-ppg conn'd. to RCS" 9331-70 Revise TS related to 5.2.2 9331-15, 10, 04 specific boron concentration in S/D margin calcs. 7 of 8 . b - . , _ _ .-. . . ., - _ -
e .. , [ IFI Number Descriptor Repcrt Section Common Items (IFI,URI,VI0s,etc) 9331-71 Revise TS related to surv. 5.2.3 9331-50, 43, 15 testing of TDAFW pmp. 9331-72 Add'l . information related 5.2.4 9331-30, 81 to GL 93-04 (Control Rod Information and Single Failure) 9331-73 Operating staff workloads 5.3.1 9331-67, 65, 60, 57, 56, 38, 37, 36, 27, 25, 06, 04. 01 9331-74 NRC assess licensee's 5.3.2 9331-44, 45, 40, 20, 21, 30 analysis of the essen. CHW sys. chiller under low heat load accident 9331-75 Licensee's corrective 5.3.3 9331-22, 17, 58 actions on fire protection deficiencies 9331-76 Tornado damper issues 5.3.4 9331-30, 40 9331-77 MOV operability / reliability 2.3.1 9331-19, 48; 9306-07 issues 5.1.12 9331-78 Speakout/EAP not anonymous 2.4.3.1 9331-56, 82 9331-79 Work procedures in error 2.2.6.2 9331-06, 07, 14, 15, 23, 56, 68 9331-80 Management support to Mgt. 2.2.5.1 9331-06, 24, 32, 49, 56, 61, 67 9331-81 Engineering backlogs large 2.3.3.1 9331-30, 31, 40, 44, 45, 46, 48, 52, 72, 74, 77 9331-82 Management response to 2.4.5.1 9331-06, 32, 53, 54, 55, 56 self-assessment functions 8 of 8 . ( b _,- . _ . , _. _ . ..,. _ . . , , }}