ML20207R998
| ML20207R998 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/28/1987 |
| From: | Julie Ward SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| JEW-87-330, NUDOCS 8703180398 | |
| Download: ML20207R998 (468) | |
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"gguy, 9pu 2 1830,(916)452-3211 SACRAMEN10 MUNICIPAL UTILITY DISTRICTSYSTEM AN ELECTRIC C 6201SERVING S Street,THE P.O.
REBox 15830, Sacramento CA 9
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iA i: 5, February 28, 1987 ,g" f 0. . ' ~, 5 JEW 87-330 Mr. J. B. Martin, Regional Administrator Regional V Office Inspection and Enforcenent U. S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596
Dear Mr. Martin:
SUBJECT:
TRANSMITTAL OF AMENDMENT 3 OF THE ACTION PLAN Transmittal is hereby made of Amendment 3 of the Rancho Seco Action Plan for Performance Improvement. For your convenience, this amendment is a complete reissue of the original document submitted to the NRC on July 3, 1986, and amended September 15, 1986 and December 15, 1986. The
" change bars" identify the changes from Amendment 2.
Amendment 3, in similar fashion to Amendment 2, identifies specific actions required for accomplishment prior to and during restart. Sections 4B and 4C, Management Actions and System Related Plant Modifications are revised and updated. l A summary of other changes incorporated in Amendment 3 appears below: The current organizational structure, which focuses on the District's restart activities at Rancho Seco, is described. The transition process and our target nuclear organization projected to exist at restart is detailed. The District's updated responses to NUREG 1195 are listed in Appendix B. The Long Range Scheduling Program is described. This program will prioritize and schedule our identified near-term and long-term (Priority 2 and
- 3) items.
8703180398 870228 PDR ADOCK 05000312 PDR j p m g
o Mr. Martin February 28, 1987
- All management effectiveness actions are listed in Section 4.B. This includes Priority 2 and 3.
- The commitments made in the previous amendments of this plan have been cross-indexed to correspond to the same issue in this amendment and are listed in Appendix D.
Specific actions to be done prior to restart are drawn together in Amendment 3. These include some items that have been designated by our Plant Performance and Management Improvement Process as Priority 2 or 3, but have been approved or designated to be accomplished prior to restart. Our priority 2 and 3 items are being scheduled in accordance with our Long Range Schedule Program. A subsequent submittal of a new Action Plan Appendix will list the Priority 2 and 3 items. Sincerely, WE E,c ( John E. Ward Deputy General Manager, Nuclear Enclosure cc: NRC, Rancho Seco ( l l
1
;fy'EO RANCHO SECO fi gj ACTION PLAN AMEN 0 MENT 3 FEBRUARY 28, 1987 g7 Old -5 'y, I
SUMMARY
OF CHANGES " ; 5j l L' ~
.i3 Page Change Reason for Change 1-13 1.25 Added Long Range Schedule Program 2-9 Figure 2-1 Reflects organizational changes.
2-10 2.2.3 New plan for transition staffing. 2-11 Figure 2-2 Reflects organizational changes. 4B Entire section Added QTS numbers where "later" appeared. 48-2 48.1.1-3 Added priority 2 & 3 items. 4C Entire section Miscellaneous corrections. 4C.1-14 4C.1.33-40 Additions to ICS/NNI work list. 4C.50 4C.50-57 New sections added. Appendix B Entire Reflects District's updated responses. O- Appendix 0 Entire Reflects cross-reference to Amendment 3. 1 0
l I INDEX
) Section Page 1.0 Introduction & Program Overview 1 .1 Post Event Review 1-2
.2 The Objectives of the Action Plan 1-2 j 1.1 Input Process (Issue Identification) 1-6
.1 Department Managers Hardware and Programmatic. 1-6 Recommendations
.2 Management Process Review 1-8
.3 Systematic Assessment Process (QCI-12) 1-8
.4 System Review and Test Program (SRTP) 1-9 1.2 Issue Evaluation and Disposition Process Management 1-10
.1 Process Overview 1-10
.2 The Recommendation, Review and Resolution Board (RRRB) 1-10
.3 The PAG Membership 1-11 x .4 Activity Prioritization Criteria 1-11 1.3 Implementation of Action Plan Activities 1-11
.1 Management, Operations, and Administrative Process 1-13 Improvement Actions
.2 Modifications and Maintenance Improvement Actions 1-13 1.4 System Review and Test Program 1-14
.1 Systems Review and Test Program 1-14
.2 System Functional Review 1-14
,- 1.5 Issue Resolution Closecut 1-16 1.6 Independent Program Oversight 1-17 1.7 Document Purpose 1-18
.1 Organization - Body 1-18
.2 Organization - Appendices 1-18 1.8 Conclusion 1-20 w, - - - . - , - , - . - ..,.-.-y . _ - . _ , _ , , - . - - _ . - , , , - , _~ ,,,r --.__,,7-__,,--..,,m,w,, , , , - , . . .y-.m.,-w . mm .. - _ - - , -
Section Page 2.0 Management of the Action Plan 2-1 2.1 Independent Review Group (IRG) 2-2
.1 Purpose 2-2
.2 Mission 2-2
.3 Tasks 2-2
.4 IRG Findings 2-3 2.2 Restart and Implementation Organization 2-4
.1 Responsibilities 2-4 The Deputy General Manager, Nuclear 2-4 The Restart / Implementation Manager (RIM) 2-4 The Implementation Manager 2-4 The System Review and Test Program Director 2-5 The Plant Manager 2-5 The Nuclear Engineering Manager 2-5 The Executive Assistant 2-5 The Quality Assurance Manager 2-5 The Support Services Manager 2-5 The Training Manager 2-5 The Licensing Manager 2-6 The Plant Modifications Manager 2-6
.2 Qualifications 2-6 2.3 Action Plan Activity Management Process 2-11
.1 The establishment of the Recommendation Review and 2-11 Resolution Board (RRRB)
.1 Criteria for prioritizing restart schedule 2-12
.2 The establishment of the Performance Analysis Group 2-12
.3 Implementation and Close-out 2-13
.4 Action Plan Activity Tracking and Reporting 2-13
. Section Page
\ 2.4 Review Meetings and Reports 2-15 Prior to Power operation Restart Report 2-15
.1 Internal 2-15
.2 External 2-15 2.5 Transition Actions 2-16 3.0 Performance Improvements Underway Prior to the 12/26 Event 3-1 3.1 Projects Underway Prior to December 26, 1985 3-3
.1 Staffing and Organization 3-3
.2 Training Program 3-3
- 1. Management Restructure 3-3
- 2. INPO Accreditation Effort 3-4
.3 Maintenance Program 3-4
.4 Quality Assurance 3-5
.5 Systematic Troubleshooting 3-5
.6 Root Cause Program 3-6
.7 Activity Assessment 3-6 4.0 Restart and Performance Improvement Action Plan 4-1 4A Systematic Assessment Program 4A-1 4A.1 Precusor Review Program 4A-2 4A.2 Deterministic Failure Consequence Analysis 4A-4 4A.3 B&W Owners Group Programs - Safety and Performance 4A-6 Improvement Program (SPIP) 4A.4 Plant Interviews 4A-9 48 Hanagement, Operations, and Administrative Process Improvement 48-1 48.1 Management Effectiveness 48-2 48.2 Quality and Quality Assurance 48.2-1 48.3 Training 48.3-1 k
Section Page 48.4 Operations and Operating Procedures 48.4-1 48.5 Maintenance Programs and Procedures 4B.5-1 48.6 Health Physics and Radiological Controls 48.6-1 48.7 10CFR50 Appendix I Discharga Guidelines 48.7-1 48.8 Emergency Preparedness 4B.8-1 4B.9 Human Factors 48.9-1 48.10 Management Information System 48.10-1 48.11 Commitment Management 48.11-1 48.12 Configuation Management 48.12-1 48.13 Materials Management 48.13-1 4C Plant Modifications and Maintenance Improvements 4C.1-1 4C.1 Integrated Control System. Non-Nuclear Instrumentation 4C.1-1 and Interfacing Systems (ICS/NNI) 4C.2 Auxiliary Steam System (ASC) 4C.2-1 4C.3 Emergency Feedwater Initiation and Control (EFIC) 4C.3-1 4C.4 120VAC Electrical System (120 VAC) Vital /Non-Vital 4C.4-1 4C.S Containment Building Spray System (CBS) 4C.5-1 4C.6 Borated Hater System (BHS) 4C.6-1 4C.7 Core Flood System (CFS) 4C.7-1 4C.8 Radiation Detector Monitor System (RDM) 4C.8-1 4C.9 Safety Features (SFS) 4C.9-1 4C.10 Reactor Protective (RPS) 4C.10-1 4C.ll Plant Security (PSS) 4C.11-1 4C.12 Nuclear Instrumentation (NIS) 4C.12-1 4C.13 Annunciator (ANS) 4C.13-1 4C.14 Plant Communication (CSP) 4C.14-1 4C.lS Computer System (PCS) 4C.15-1
3 Section Page ( ,,
,A 4C.16 Nuclear Service Water (NSW) 4C.16-1 4C.17 Nuclear Raw Water (NRW) 4C.17-1 4C.18 4160VAC 4C.18-1 4C.19 -Lube 011 System (LOS) 4C.19-1 4C.20 Purification and Letdown (PLS) 4C.20-1 4C.21 Decay Heat (DHS) 4C.21-1 4C.22 Reactor Coolant System (RCS) 4C.22-1 4C.23 6900VAC 4C.23-1 4C.24 Main Feedwater (MFW) 4C.24-1 4C.25 Main Steam System (MSS) 4C.25-1 4C.26 Gland Steam and Condensate (GSC) 4C.26-1 4C.27 125VDC Vital & Non-Vital 4C.27-1 T
4C.28 480VAC 4C.28-1 4C.29 Main Generator Seal Oil (MGS/ SOS) 4C.29-1 4C.30 Once-Through Steam Generator (OTSG) 4C.30-1 4C.31 Site Reservoir System (SRS) 4C.31-1 4C.32 Instrument Air System (IAS) 4C.32-1 4C.33 Aux 111ary Feedwater (FWS) 4C.33-1 4C.34 Essential HVAC, Control Room and NSEB 4C.34-1 4C.35 Reactor Coolant Drain (RCD) 4C.35-1 4C.36 Drainage and Sewerage (CDS) 4C.36-1 4C.37 Radwaste (RHS) 4C.37-1 4C.38 Control Rod Drive (CRD) 4C.38-1 4C.39 Waste Gas System (WGS) 4C.39-1 4C.40 BOP HVAC 4C.40-1 4C.41 Reactor Sample System (RSS) 4C.41-1 Section Page 4C.42 Emergency Generator Diesel Fuel Oil System (EGS) 4C.42-1 4C.43 Seal Injection and Make-Up (SIM) 4C.43-1 4C.44 Carbon Dioxide System (CO2) 4C.44-1 4C.45 Fire Protection System (FPS) 4C.45-1 4C.46 Main Condensate System (MCM) 4C.46-1 4C.47 Plant Buildings and Structures (PBS) 4C.47-1 4C.48 Control Room /TSC Ess HVAC (HVS) 4C.48-1 l 4C.49 Component Cooling Water (CCH) 4C.49-1 4C.50 Main Control Room (MCR) 4C.50-1 1 4C.51 Heaters, Drains and Vents System (HDV) 4C.51-1 4C.52 Main Turbine and Extraction Steam System (HPT) 4C.52-1 4C.53 Main Turbine Electro-Hydraulic Control System (EHO) 4C.53-1 4C.54 Secondary Chemical Addition System (SCA) 4C.54-1 4C.55 Plant Cooling Water System (PCW) 4C.55-1 4C.56 Nitrogen Gas System (NGS) 4C.56-1 l 4C.57 Hydrogen Gas System (HGS) 4C.57-1 4D System Review and Tasting Program 40-1
- 1. Purpose 40-1
- 2. Organization 40-2
- 3. Responsibilities 40-5
- a. System Review and Test Program Director 40-5
- b. Test Review Group 40-5
- c. Performance Analysis Group 40-6
- 4. System Selection for Systems Review 40-7 O
Section Page
- 5. System Review Report Summary 40-10
- a. System Status Report Overview 40-10
- b. System Investigation Report Overview 40-10
- c. Details of Report Components 40-11
- 6. Restart Test Program 40-14 40.7 Comparison Between Rancho Seco and Davis-Besse Programs 4D-16
.1 System Selection 4D-16
.2 Problem Identification 40-17
.3 System Review 40-18
.4 Test Program Development 40-19
.5 Restart Testing 40-20 APPENDIX A A-1 District Board of Directors' Policy Statement on Performance A-1 O Improvement at Rancho Seco APPENDIX B B-1 Specific District Responses to NUREG--Il95 Findings B-1 Findings 8-1 B.1 ICC/NNI Power Supply Monitor B-1 B.2 Repositioning of ICS Controlled Valves or Loss of ICS B-2
- B.3 PM Program for Manual Valves B-2 B.4 Procedural Guidance for Loss of ICS B-3 B.5 Feedpump Trip Criteria B-4 8.6 Priority of PTS or Pressurizer Level B-4 B.7 Training on Loss / Restoration of ICS Power B-7 B.8 Recognition of ICS Power Condition B-9 i
l I
Section Page B.9 Damaged Hand Operator on AFH Valve B-10 B.10 Radiological Controls and Emergency Preparedness B-10 B.11 Installation of EFIC B-11 B.12 Reactor Vessel Thermal Shock B-12 B.13 Reactor Vessel Integrity B-13 B.14 Use of ICS in FSAR Design Basis Events B-13 B.15 Precursors to 12-26 Event B-15 B.15.a Power Supplies to ICS/NNI B-15 B.15.b January 5, 1979 Loss of ICS Power B-16 B.15.c ICS Rella.bility Study, BAW-1564 B-16 B.15.d District Response to IE Bulletin 79-27 B-24 B.15.e District Response to February 1980 Loss of NNI at B-25 Crystal River B.15.f District Response to NUREG-0667 B-26 B.15.g Significance of Partial loss of NNI at Rancho Seco B-26 March 1984 B.15.h Appilcability of " Reference Plant" studies to B-27 Rancho Seco B.16 Timely Identification of Loss of ICS Power Condition B-27 B.17 Usefulness of Annunciator Procedures Manual B-28 8.18 Performance of ICS Following Restorat on of Power B-28 B.19 Control Room Indicators Hhich Fall to Mid-Scale B-29 B.20 Adherance to Radiation Protection Requirements B-30 B.21 Programmatic Efforts to Disseminate Lessons-Learned and B-30 Plant Changes B.22 Operating Crew Minimum Required Staffing B-30 B.23 Role of STA B-31 B.24 Application of Systematic Troubleshooting B-31 0
!- 3 4
t Section ' Pace . i B.25 IIT Requests for Information B-32 1 B.26 Applicability of Generic PTS Analysis to Rancho Seco B-32 4 i . i APPENDIX D i , 1 Action Plan Commitment Summary D-1 i a j ! I APPEN0!X E 5 1 Sample Portion of Action Plan Activity Tracking Report E-1 , i i APPEN0!X H i, l Sample Test Speelfications H-1 , { i APPEN0!X I j l Plant Staff Interview Program I-I , l (Excerpt from QCI-12) l l l i i 1-I i } ' l. 1 4 - I l l l i h I I i I I t i , I
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INDEX OF FIGilRES NA NAME PAGE O 1-1 Post-Trip Window 1-3 1-2 Issue Evaluation and Disposition Process 1-5 1-3 Identifying the Issues and Problems Associated 1-7 with Plant Operating Performance 2-1 Restart / Implementation Organization 2-10 2-2 Level One Ultimate Organization 2-11 4A-1 B&W Owners Group Safety and Performance 4A-8 Improvement Program 40-1 System Review and Test Program Organization 4C-3 4D-2 System Reviews and Reporting 40-4 40-3 System Review and Reporting 40-16 O O
1.0 INTRODUCTION
& PROGRAM OVERVIEW The December 26, 1985 overcooling incident at Rancho Seco has prompted a comprehensive investigation that looks far beyond the specific problems directly associated with that incident. The steadily degrading performance of the plant and its staff are symptomatic of more serious deficiencies than those associated with the December incident.
While plants stellar to Rancho Seco have achieved performance levels consistent with the better performers in the industry, the operating record of the Rancho Seco Nuclear Plant, as measured by plant performance statistics (capacity factor, etc.), Systematic Assessments of Licensee Performance (SALP), and INPO performance Indicators has not been satisfactory. Subsequent to a 1984 evaluation by consultant LRS, the Sacramento Municipal Utility District (SMJ0) Board of Directors decided to take action to improve the Rancho Seco performance. To achieve this objective each of the areas affecting plant performance was investigated or studied. These areas included plant hardware, management and administrative systems, organizational structure and staffing, maintenance, training, personnel, and physical facilities. To implement and achieve observable results from the changes indicated by these studies requires time and significant financial commitment. The District Board of Directors decided in 1984 that the existing investment, combined with the projected electric demands in the SMUD service area, and the benefits O to be derived from achieving a higher level of performance, justify the additional investment required to achieve the designed results. They also realized that reliability improvement is closely coupled with safety improvement which has always been a first priority. 8efore actions associated with this decision to improve performance could take effect, the importance of the program was reinforced by a number of undesirable operating expertences in 1985. The most significant of these events occurred on December 26, 1985, when a loss of power to the plant's integrated control system led to a plant overcooling. The cooldown rate specified to Ilmit the stresses induced in reactor systems heavy metal components was exceeded. Wille subsequent analyses determined that no serious stresses were induced, the significant potential of this event is not to be understated. Following the event, the District and the NRC Independently conducted reviews to determine the nature and extent to which management, programmatic and hardware deficiencies contributed to this--and previous--incidents. The District has documented its findings in the IAG Root Cause Report 85-41 and the NRC Incident Investigation Team has documented their findings in NUREG-1195. The conclusions and recomendations identifled and contained in NUREG-1195 are consistent with those reached by the District. In general, these findings are: O 1-1
- a. The trip and the associated rapid cooldown was caused by the failure of Rancho Seco to implement design changes in a timely manner which would have compensated for known design weaknesses,
- b. The failure of Rancho Seco to implement adequate compensatory measures for the design weakness, such as procedural guidance and training, contributed to the significance of the event.
- c. Maintenance program deficiencies contributed to the inability to mitigate the severity of the cooldown transient.
- d. Non-compliance with existing precedures contributed to the overcooling event and caused additional unnecessary complications,
- e. Manufactaring defects in the electrical terminations of particular control cabinets (ICS) initiated the event.
1.0.1 Post Event Review The post event reviews by the Olstrict and the NRC identify the specific actions to minimize the potential of this event occurring again and those necessary to assure the event did not degrade or impact the ability of the plant to operate safely and reliably. These pertinent documents and actions include the following: IAG Root Cause Report NUREG 1195 Equipment Failure Investigations
- ICS
- ICS Controlled Valves
- Makeup pump
- Radiation Monitor Effects of overcooling transient on components Operational Review (including adequacy of procedures)
Human Factors Evaluation Adequacy of Training Thermal / Hydraulic Response of the Reactor Coolant System Health Physics Emergency Preparedness Based on these findings, SMUD has modified, expanded, and accelerated the implementation of its performance improvement action plan. 1.0.2 The objectives of the Action Plan are to:
- 1. Reduce Reactor trips
- 2. Reduce challenges to safety systems
- 3. Assure the plant remains in the post-trip window (The allowed ranges of reactor coolant system pressures and temperatures immediately following a reactor trip, see Figure 1-1).
- 4. Assure compilance with Itcense requirements
- 5. Minimize the need for operator actions outside the control room.
- 6. Improve the reliability and availability of the plant 1-2
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The Action Plan has been structured to achieve these objectives through the implementation of a number of individual program elements. A schematic representation of these Action Plan Program elements and their relationships is contained in Figure 1-2. In general, these program elements are structured to: a) assure that issues or deficiencies in plant design, operations and operating procedures, management and management processes, training, etc., which have the potential to contribute to an event such as the December 26, 1985, event, or negatively impact the performance of the Rancho Seco power station are identifled and input to the action plan for evaluation and resolution. b) assure that each of these issues receives a thorough evaluation and is properly dispositioned. c) assure that actions are implemented in an efficient, effective, and timely manner consistent with their importance to safety and reliability of plant operations, d) assure that closure of the action items is complete, addresses the issue adequately and that the actions are taken in accordance with the approved plant procedures. O O 1-4
i ISSUE EVALUATICN & DISPOSITION l l PROCESS NPUT MANAGBfNT FPLEffNTATION CLOSURE i Dept. Tigrs. Restart Testing ] Hardware & F1odifications & j Programmatic 3 t1aintenance j Recommendations Improvements 1r flanagement Operations &
'I 1' Administrative Verification
! a T1anagement Performance Process Validation j Process : Analysis Improvements Filing j Review Group 4 a i 1 " l - Tracking I, Systematic SYSTEt1S REVIEW Recommendations
- Assessment -
' Review and ; #8 .
s Wr. 1 Functional Regnts. Resolution Board Test Reqmts. 1 I-12) 1 l i i Figure 1-2 1 8
1.1 INPUT PROCESS (ISSUE IDENTIFICATION) The process is designed to assure that the review of the various areas which can impact plant performance (i.e., management and management processes, plant design, operations and operating procedures, maintenance, training, and other support activities) is adequate to identify any deficiencies. The process consists of activities which review these impact areas from four perspectives, 111ustrated in Figure 1-3: (1) the top down department managers hardware and programmatic recommendations; (2) the management process review; (3) the bottom-up systematic assessment program elements; (4) and the system review and test program. Each of these perspectives has advantages and disadvantages relative to its effectiveness and efficiency in identifying deficiencies and developing improvement actions to be taken. By incorporating key features of each, we have tailored an action plan which is diverse and broad in scope, directly addressing the type of deficiency which has contributed to the poor performance record at Rancho Seco and the December 26, 1985, event. A description of these Issue Identification Action Plan elements is as follows: 1.1.1 Department Managers Hardware and programmatic Recommendations I An assessment of the plant design, management, operations, and administrative system deficiencies was conducted based on the functional organizations' knowledge and existing documents of previous evaluations by others. The following important sources were used as input to this assessment. O l-6
e Mgure 13 IDENTIFYING THE ISSUES AND PROBLEMS ASSOCIATED WITH PLANT OPERATING PERFORMANCE DER (RrMENrMAN G H ett and Programatic 2"a s",'u" N
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T LRS Management Audit INP0 Audit Reports Commitment Lists American Nuclear Insurers (ANI) Open Items NUREG 1195 IAG Report 85-41 This assessment led to the development of the management, operations, and administrative process improvement action plans (see Section 48) and the plant modification action plans. This effort was known as the " Top-Down" determination of restart requirements. With completion of the Systematic Assessment Report Program, the
" Bottom-Up" listing of restart items became available. The Action Plan provides the listing of restart items. These are listed in Section 48, Management, Operations, and Administrative Program Improvement, and in Section 4C, Plant Modifications and Maintenance Improvements.
1.1.2 Management Process Review The management process review was conducted by the management process review group which has:
- a. Reviewed previous management audits and assessments from the last five years to date.
- b. Reviewed SMUD responses / commitments to these documents,
- c. Conducted a direct status assessment of current management processes and functions,
- d. Abstracted management assessments from the other Plant Performance and Management Improvement Program (PP&MIP) investigations.
The plans to improve management effectiveness which resulted from these actions are contained in the management, operations, and administrative process improvement action plans (see Section 48). The MPRG team will facilitate the implementation of all management effectiveness action plans. 1.1.3 Systematic Assessment Process (OCI-12) The Plant Performance and Management Improvement Program (PP&MIP) is a comprehensive broad-scope, systematic assessment program. This program was developed and is being Implemented to perform a detailed review of the plant design and expertence as well as the appropriate Industry experience. This program will confirm the action plan elements based on the department managers' assessment, and identify any necessary enhancements to these plans to address any deficiencies identified through this program. Specific input areas include:
- a. Precursor Reviews
- b. Plant Staff Interviews
- c. Deterministic Failure Consequence Reviews 1-8
l
- d. B&W Owners Group Safety and Performance Improvement Program O (SPIP)
\'j e. December 26 Event and NUREG-11951tems i f. Selected projects This process develops recommendations, which must be resolved, and provides a structured program with prioritization criteria to ensure
, that each item is appropriately resolved. The actual disposition of I a recommendation will be done uttilzing the established i administrative and programmatic processes in place within the nuclear l organization. As an example, a recommendation, which when l dispositioned, requires a modification to the Control Room will, as a ! part of the configuration control and modification programs, receive c Human Factors review against the CRDR criteria and commitments. In addition, it will be reviewed by Operations for E0P and Training Impacts, and may well require that a Verification and Validation (V&V) exercise be performed prior to incorporation. The Action Plan adds no steps to these programs, it does add significantly to the number of items being processed through these programs. 1.1.4 System Review and Test Program (SRTP) The system review and test program is a key element in the implementation process. The systens engineers implementing this program are in a position to obtain an overview perspective of the issues and recommendations. This overview perspective combined with the implementation of the system review process may lead to the l Identification of additional Issues and recommendations. The details of this comprehensive System Review and Test Program are provided in Section 40, which includes definition of the role of the System Engineers, l O V l-9
1.2 ISSUE EVALUATION AND DISPOSITION PROCESS MANAGEMENT The process to address and manage the resolution of issues is described in detail in a special Plant Procedure, QCI-12. An overview of the steps of this process as well as a description of the RRRB, PAG, the prioritization process and the oversight role of the independent review group is as follows. The Plant Performance and Management Improvement Programs (PP&MIP) as defined within QCI-12, is intended as a cae-time ! effort, although specific elements will be incorporated into the administrative and operational processes. The program identiftes issues and provides for their prompt review and disposition. The disposition implementation is accompItshed by uttitzing the existing administrative processes procedures, and programs. While these are themselves subjects for the PP&MIP, they are not supplanted by the PP&MIP. As examples, the Plant Review Committee (PRC) continues to review and approve procedures, while the Management Safety Review Committee (MSRC) performs its duties in addressing nuclear safety concerns and issues. Items involving safety issues, whether initiated by the PP&MIP or otherwise, are addressed and resolved before these committees. 1.2.1 Process Overview Issues, along with recommended solutions, are identified by the special task or input groups. These issues and recommendations are sent to the Review, Recommendation, and Resolution Board (RRRB) for validation and acceptance. Recommendations involving systems under review are sent to the System Engineer and others are sent directly to the Performance Analysis Group (PAC) (made up of the Nuclear Department Managers). All recommendations, both valid and invalid, are dispositioned to the satisfaction of the PAG. This group confirms the disposition of the priorttles and valid issues and sends them on to the appropriate department for implementation. The invalid issues and recommendations are sent to QA for independent review and formal close out and for system related items, to the System Engineer for information. Once implementation action on valid recommendations is completed a close out record is prepared and sent to QA for vertftcation of content and validation that the implemented action is consistent with intent of the issue and recommendation. 1.2.2 The _ Recommendation. Review, and Resolution Board (RRRB) is a nine member multidisciplined group oIlndividuaTs with nuclear expertence and training drawn frcm SHUD, another utility wtth a B&W NSSS, NSSS Vendor, and the plant Architect Engineering firm. O I-10
I As described in QCI-12, the RRRB is to determine the " validity" i.e., the correctness and uniqueness, of each received recommendation. s i They consider only the technical merits, not the cost, time, or resources available. Once they process a recommendation, it is passed to the System Engineer or the Performance Analysis Group (PAG) for evaluation and disposition. 1.2.3 The PAG membership is composed of the Managers of the Nuclear Departments or their designees. This group has knowledge of the competing priorttles, needs, commitments, and resources available to resolve each recommended action. The member nuclear departments are: Executive Assistant / Manager, Nuclear Projects, Chairman Licensing Quality Plant Nuclear Engineering Training In addition to receiving input from the RRRB, the PAG also receives inputs from the Management Process Review Group, the Systems Engineers for System Review and Test issues, and recommendations from the department managers themselves. All input is evaluated against i the objectives of the program as defined previously in this section. The PAG then determines the disposition and assigns the appropriate priority to each item. i pl Implementation of the dispositioned recommendations after approval by the Deputy General Manager, Nuclear, falls to the line organization (d which utilizes the processes and controls governing these activities. This includes resource allocation and scheduling of plant affecting work, configuration control processes, training programs, and i operations, 1.2.4 Activity Prioritization Criteria The criteria by which activities to resolve issues are prioritized and placed in the appropriate disposition period is as follows: Restart - Actions to be completed prior to restart or completion of the Restart Test program which will: . a. assure the plant remains within the post-trip l window
- b. assure compItance with technical specifications
- c. minimize the need for Operator Action outside the control room within the first ten minutes of an i event.
Near Term - Actions to be initiated as promptly as practicable, schedule developed, resources assigned and maintained until completed which will:
- a. enhance the ability to remain in the post trip window (e.g., auto action vs. operator action) 1-11
- b. reduce reactor trips
- c. reduce challenges to safety systems
- d. produce near-term programmatic benefits.
Long Term - Actions to be programmed for the longer term which will support the achievement of the 1990 INP0 plant perfor mance objectives, i .e.
- a. Improve reliability
- b. Improve availability
- c. major programmatic enhancements 1.2.5 Long Range Scheduling Program The near term and long term schedule program is being developed to provide a control tool for the identification, evaluation, budgeting and scheduling, implementation, and monitoring of the implementation program. The program design is intended to satisfy multiple needs at Rancho Seco. Among these needs are providing a tool for planning and tracking commitments established by the Business Plan, the Living Schedule, long range departmental goals and objectives as well as the longer termed performance improvement program.
This Long Range Scheduling Program (LRSP) is a continuous process of selecting, Integrating, prioritizing, and scheduling plant and organizational betterment activities. The selection of activities is based on safety, regulatory, reliability, operability and economic factors. These activities are prioritized and scheduled in order to optimize the allocation of resources for assuring the protection of public health and safety and for the continuous upgrade in plant and personnel performance. The key process elements which form the basis of Rancho Seco's LRSP are:
- 1. Issue identification and evaluation, including scope definition.
- 2. Issue priorttization in accordance with established crtteria based on resource availability, technical reviews and regulatory requirements.
- 3. Identification and evaluation of alternatives.
- 4. Final design and evaluation of selected project methodology.
- 5. Prioritization and categorization of projects in accordance with defensible analyses.
- 6. Development of Integrated Implementation Schedules.
- 7. Nuclear Regulatory Commission agreement, where appropriate.
O l-12
4
- 8. Implementation of activities to effect the change.
( 9. Performance monitoring.
- 10. Annual re-evaluation of the LRSP.
This process is being initiated at Rancho Seco during the next two , months. The initial phase will be to procedura11ze the process and staff the organization. This process will then address the identified Priority 2 and 3 items from the PP&MIP for cost and , prioritization evaluation and development of detailed schedule plans. The present target is to have the Priority 2 and 3 items prioritized and scheduled prior to Restart. It is our intent to list 1 these Priority 2 and 3 items in a separate submittal or in an Appendix to the Action Plan at a later date. 1 ( 5 i 1 ,f i i i I 4 i j 1-13 3
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l l 1.3 IMPLEMENTATION OF ACTION PLAN ACTIVITIES The majority of the actions to evaluate the impact of the December 26, 1985 event and to preclude this event from recurring O1 have been completed by the District with many receiving concurrence of closure by the NRC. The implementation of additional actions to address the broader performance improvement issues have been identified by the District's Nuclear Department Managers. In order to implement and complete Action Plan items and other commitments, a comprehensive tracking system has been developed in accordance with QCI-12. The Action Plan describes action scheduled for restart cross-referenced Tracking System (QTS), when applicable. The Action Plan is not intended to document all actions to be completed prior to restart. The two areas of implementation are; a) management, operations, and administrative process improvement actions and b) plant maintenance and modification actions. Each of these areas is described below. 1.3.1 Management, Operations, and Administrative Process Improvement Actions The major management and administrative process is described in Section 48. The QCI-12 Tracking System, typified by Appendix E, lists the valid programmatic improvement actions associated with management, operations, and the administrative processes. This portion of the Action Plan has been divided into sections for each management area and subdivided into specific action items identified as necessary for accomplishment prior to restart. Specific closure requirements have been detailed and projected completion dates are provided. 1.3.2 Modifications and Maintenance Improvement Actions A description of the specific major modifications and maintenance improvements is provided in Section 4C. Section 4C describes items scheduled for completion prior to restart. Derived from the system reports, Section 4C subsections describe all the problems associated with each system, their closure requirements, and projected completion date. Most of these items include more than a single recommendation or disposition to accomplish their closure. Appendix E lists examples of the valid recommendations being developed and processed. The entire QCI-12 Tracking System (QTS) list is available at Rancho Seco. It is expected that some new items will continue to be identified and processed through the closecut of the Systematic Assessment Program. O 1-14
i
- 1 . -i LE
!- 1.4 SYSTEM REVIEW AND TEST PROGRAM
((1* There are two system review programs to be. conducted as part of this action plan. f!- The first program is modeled after the Davis-Besse program and is a key element in the restart of-Rancho Seco. This program is structured to provide a systems review of issues, and recommendations as well as to develop'a restart test program. This action will provide additional assurance that these systems have retained their FSAR/USAR functional basis, or have adequate analysis to justify differences, and that they have been adequately tested. The second program is a longer term program modeled after the NRC's Safety System Functional Inspection. This program provides a more
;- detailed-look at the reliability and component design criteria.
j 1.4.1 System Review and Test Program [ The System Review and Test Program (SRTP), which is modeled after the Davis-Besse program, is a key element in the implementation process. j It consolidates all system related issues with the relevant
- . recommendations from the Systematic Assessment Program, for coordinated resolution. This consolidation is done by an assigned System Engineer who performs supplementary tasks such as walkdowns, i functional criteria development, review of other corrective action i
systems for system issues needing resolution prior to restart, and j - develops an integrated program for resolution of system issues. The l~ systems engineer then defines any necessary testing and functions as ! the test engineer to accomplish it. The result is a determination that the system is ready for plant operation. l The systems-engineer program being implemented at. Rancho Seco is i modeled after the INPO G000 PRACTICE . In this program'the system - engineer is responsible, among other things for: (1) the development i of system functional requirements; (2) assuring coordinated and i effective disposition of system deficiencies; (3) assure the adequacy i- of system testing; and (4) to develop and maintain a set of test i requirements which assure the material readiness and operability of j each system. A selection criteria has been established to divide the major systems into two categories, " Selected" and " Additional". In both cases the system functional criteria are developed in combination with the problem statements and presented to the Performance Analysis Group ,
- i. (PAG). Pending their concurrence, the recommendations of the System i Engineer are implemented, and those systems in the " Selected" categories are tested (prior to or during) the restart.
1.4.2 System Functional Review A review of reports on Reactor Trips at B&W plants was performed as part of the Precursor Review, an element of QCI-12. This review showed that over 40% of these transients were due to mismatches between reactor heat generation and secondary side heat removal. 1-15 p
. W= ww w Wye . *esey r t= w"& .
Consequently, systems critical to secondary side heat removal will undergo a long-term and more extensive system review than that described above. This process, which is modeled after the safety system functional inspections by the NRC, consists of: (a) Design Basis reconstitution; (b) reliability assessment of the system; and (c) the evaluation of individual component design criteria to assure that the individual components support the system design basis. A reliability assessment of these systems will also be conducted as part of this program to determine which components of these systems are critical to the prevention of reactor trips and which are critical to assure that, immediately after a reactor trip, the transient remains in the post-trip window (see Figure 1-1). The system surveillance tests, where appropriate, will be evaluated to assure they adequately demonstrate the operability of the system and/or components to meet their design basis requirements. The five systems selected for this comprehensive review are: .1 Main Feedwater System .2 Auxillary Feedwater System .3 ICS/NNI .4 Pressure Control functions of the Main Steam System .5 Instrument Air This comprehensive review of these five selected systems will be initiated prior to restart and will be completed prior to coming out of the cycle 8 refueling outage. O l-16
1.5 ISSUE RESOLUTION CLOSE00T A formal process will be implemented to assure the effective and
- complete closeout of action plan items. This is accomplished through the Quality Assurance department which is charged with the responsibility to verify that actions were taken in accordance with the plan and existing procedures and to validate that the actions taken meet the intent of the original recommendation and resolve the original issue.
0 4 4 4 i i i i l 1-17 l
~
l l__ _ _ ___ _ _, . _ . . _ . _ _ _ . _ . _ _ . _ . _ _ . _ . . . _ _ . _ . _ _ _ _ _ , . _ _ _ _ _ . _ _ . . . _ _ . _ _
l.6 INDEPENDENT PROGRAM OVERSIGHT Independent oversight of the Action Plan is provided by an Independent Review Group (IRG). This group consists of senior persons with significant experience in the management and oversight of nuclear power plant operations, design, and regulations. O O 1-18
1.7 DOCUMENT PURPOSE f Nj The purpose of this document is to communicate the District's planned actions and program status, both internally and externally. The body of the document contains a description of the program elements and a general description of the actions to be taken. The document's appendices provide descriptions of the detailed actions to be taken, and their status, scheduler information and supplemental information such as responses to the NUREG-1195 conclusions and recommendations. This particular structure was selected to accommodate the unique features of the District's program. 1.7.1 Organization - Body The body of the report is organized as follows:
- a. Section 2.0 Management of the Action Plan - describes management processes and features of the program to assure that appropriate actions are being taken to effectively, efficiently, and thoroughly identify, prioritize, control, and implement changes to resolve the type of deficiencies which contributed to the December 26, 1985 event and the poor performance record of Rancho Seco. This includes:
The Independent Review Group The Restart and Implementation Organization (RIO) The Action Plan Tracking, Reporting and Close Out Process p) (
%./
The Action Plan Adjustment Process The Transition Plan
- b. Section 3.0 Improvements Prior to the December 26, 1985 Event -
describes those actions which the District had taken prior to the December 26, 1985 event.
- c. Section 4.0 Action Plan - provides a description of the actions to be taken to address the concerns raised by the December 26, 1985 event and those associated with the poor performance record i of Rancho Seco. This information is contained in the following subsections.
- 1) The systematic assessment process elements.
- 2) The management, operations, and administrative process improvement actions.
- 3) The plant modifications Actions.
- 4) The System Review and Test Program.
1.7.2 Organization - Appendices The appendices of the report are organized as follows:
- a. District Board of Directors Performance Improvement Policy l Statement.
1-19
- b. The District's Response to the NUREG 1195 Findings and Conclusions.
- c. Deleted - Incorporated Reference of NRC Open Items in Action Plan Sections,
- d. Action Plan Commitment Summary.
- e. Sample Action Plan Activity.
- f. DeletiJ - Major Milestone Schedule.
- g. Deleted - Example System Review and Test Report (Section 4C describes).
- h. Sample Test Specifications.
- 1. Plant Staff Interview.
O O l-20
l.8 CONCLUSION
/ As the District management team shaped this program, it became clear that there was a need for policy direction regarding the long-term future of the Rancho Seco Nuclear Generating Station. This policy direction is considered essential to provide a foundation for long-term planning. The policy direction has been prepared and was unanimously endorsed by the Board of Directors. A copy of this Board Policy Statement is included as Appendix A.
In summary, the Rancho Seco Action Plan is an enhanced and accelerated version of & program already underway at the time of the December 26, 1985, event. This action plan is intended to re-establish a dedication to excellence which will be the basis for regaining the confidence of regulators, county and state officials, investors, and customers. The Action Plan constitutes a complete reassessment of management's role in establishing an environment in which excellent performance is expected and in which any deviation from excellence is cause for prompt and aggressive corrective action. While the number and scope of activities contained in this plan are significant, a large number of activities important to restart and performance improvement at Rancho Seco have been completed. The SMUD team is dedicated to bringing about this performance improvement in an orderly, safe, and effective manner and believes this Action Plan to be the vehicle for such change. O F n v I-21
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- - 2.0 MANAGEMENT OF THE ACTION PLAN This section describes the actions that the District has taken and i the processes which have been established to assure that the Action Plan is thorough and comprehensive, and can be implemented in an effective, efficient, and timely manner.
To accomplish this the District has taken the following steps: a) Established an Independent Review Group (IRG). b) Established a Restart and Implementation Organization (RIO). c) Established a management process to identify, track, control, implement, and close out deficiencies associated with the management and operation of the facilities, d) Developed a transition strategy for the long term which incorporates appropriate features of the Action Plan into the line organization to achieve and maintain a high level of plant performance. O 9 O 2-1
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2.1 INDEPENDENT REVIEW GROUP (IRG) 2.1.1 Purpose An IRG, made up of Messrs. John Jackson, Richard De Young, James O'Hanlon, and Arthur Gehr, has been established to periodically provide the District's General Manager and Deputy General Manager, Nuclear with assessments as to the effectiveness of the Plant Performance and Management Improvement Program, and the readiness of the plant to restart and operate safely and reliably. These four individuals, who comprise the IRG, encompass a broad range of pertinent experience. John Jackson - Many years as a quality expert in design, construction, operation and training, as well as the analysis and oversight of quality functions. Richard De Young - A distinguished career with key assignments in nuclear regulation and inspection and enforcement, as well as current experience in the analysis of nuclear management. James O'Hanlon - Significant responsibilities in plant maintenance, and operation with current oversight responsibilities at Davis Besse. Arthur Gehr - A recognized expert in nuclear law and regulation who played a significant role in the development of the nuclear programs at Commonwealth Edison and Arizona Public Service Company. 2.1.2 Mission The mission of the IRG is to assure the General Managar and the Board of Directors that the Action Plan for Performance Improvement at Rancho Seco is thoughtfully developed and appropriately implemented such that Rancho Seco is properly prepared for restart and ongoing operation with a reasonable likelihood that it will operate safely and reliably. 2.1.3 Tasks The following tasks will be performed: Interview personnel
- 1. Key SMUD managers and supervisors
- 2. NRC personnel on-site, in Region V, and Hashington
- 3. Rancho Seco staff personnel
- 4. Others Carl Andognini (Consultant to the Board)
J. Mattimoe (Former SMUD General Manager) E. Hilkinson (Former President INPO) INPO (all key managers) The SMUD Board of Directors 1 2-2 r
Attend key meetings NRC Region and Headquarters meetings Special Davis-Besse Review Group Attend Formal Presentations given by the following organizations to relate the progress in each of the functional areas over the course of the Action Plan implementation. Engineering
- Quality Assurance Purchasing Maintenance Licensing (USAR)
- Training Operations
- Regulation Review Technical Presentations IIT Report SALP Report history LRS INPO PAT Reports Participate in Plant Tours Review Organizational Issues Organization Functional responsibilities fN
- Personnel Delegation of authority Perform reviews of key operating and administrative procedures.
2.1.4 IRG FINDINGS The IRG is to make findings in four areas. Are the responses to the immediate implications of the December 26, 1985 overcooling event appropriate and complete? Is the plan to identify programmatic deficiencies and the broader implications of the plant's past history and operating performance adequate and are the planned corrective actions sufficient to reasonably expect safe and reliable operation? Are the priority of action criteria appropriate to assure that those actions taken before restart are sufficient to reasonably assure safe operation? Is the executive direction and the management plan appropriate to provide proper implementation of the Action Plan?
- v 1
2-3 u-+e e 4--e- + -memm-e , rg - w -w.p---ve r-"ew -+' * - - - "m- -"="*-a-w------a" -~-w'-----+- '* -*-'"*'T- W v-r---Tw
2.2 RESTART AND IMPLEMENTATION ORGANIZATION The Restart and Implementation Organization is a temporary organization designed to augment and assist the normal line organization in the restart and implementation of improvement action items. This organization is shown in Figure 2-1. This most recent organizational structure provides for a more efficient focus on the District restart activities while incorporating the necessary organizational elements to facilitate a smooth transition to a totally SMUD managed Rancho Seco facility. This organization will remain in effect until disbanded by the Deputy General Manager, Nuclear. 2.2.1 Responsibilities The Deputy General Manager, Nuclear provides management oversight to the implementation of the Action Plan and assures an effective interface with the normal line organization. He is responsible for approval of the Action Plan content and the schedule for implementation. He has the authority to abolish the temporary organization when, in his judgement, the additional resources are no longer required to ensure the timely implementation of the action items. The Restart / Implementation Manager (RIM) is responsible to the DGM, Nuclear for directing, through the Restart and Implementation Organization, the implementation of the restart and improvement programs. In this capacity, he is accountable for scope, schedule and incremental costs for the activities which must be completed to facilitate restart and to comply with the commitments of the action plan. The RIM has the authority to adjust schedules and resource applications within pre-approved action item categories without the AGM's prior approval to assure efficient application of resources and consistency of plant, system, and component conditions. The RIM provides functional and administrative direction to the Plant Manager. The RIM provides functional and administrative direction for restart and improvement activities to the Implementation Manager and the Nuclear Department Managers. The RIM has authorization to issue instructions as needed to further define the Restart and Implementation Organization or to control interfaces to effect the action plan. The Implementation Manager is responsible to the RIM for execution of the physical work items required for restart. The Implementation Manager also provides scheduling for the entire Restart and Implementation Organization including programmatic commitments. He has the authority to rearrange sequence of work within limits of the plant technical specification requirements, plant operating procedures and schedule commitments. Changes which would require extension of schedule commitments, or plant procedure changes, shall only be made with the concurrence of the RIM, and for the latter case, the Plant Manager. 2-4
The System Review and Test Program Of rector is responsible to the [ s Plant Manager for developing and implementing the test program identified in the action plan. This includes: identification of organization and resource needs; development of specific test objectives and acceptance criteria; development of special test procedures (as required); perform special test procedures; coordinating the performance of surveillance tests; working with the Implementation Manager in the development of a detailed test 4 schedule; and evaluaticn of test results. All procedures used in the test program shall be reviewed, approved and used in accordance with AP. 2, " Review, Approval and Maintenance Procedures", AP. 302, "Special Test Procedures", or AP. 303, " Surveillance Program". The Plant Manager Is responsible to the RIM for providing the , required operations, maintenance, health physics and technical support to accomplish the scheduled hardware and programmatic activities. This includes timely review of procedures and test results by the PRC and MSRC. He is responsible for the preparation of reports as required by the RIM and is responsible for the System Review and Test Program. The Nuclear Engineering Manager is responsible to the RIM for providing the required engineet ing support for design modifications, engineering studies and other engineering support required to accomplish the schedule hardware and programmatic activities. He is responsible for the preparation of reports as requested by the RIN. The Executive Assistant, in the role of Performance Improvement Manager, is responsible to the Deputy General Manager, Nuclear for , the activities of the Performance Analysis Group and those review activities conducted under the auspices of QCI-12. He is also responsible for preparation of reports as required by the RIM through the Deputy General Manager, Nuclear. The Quality Assurance Manager is responsible to the Deputy General Manager, Nuclear for providing appropriate surveillance and quality , engineering to assure the scheduled activities are conducted in accordance with the District's legal commitments and quality
, programs. He shall also be responsible for developing and implementing a method for verification of completion of action plan items (QCI-12). Corporate QA will retain the independent audit function.
The Support Services Manager is responsible to the RIM for providing the required support in the areas of materials management, information systems, and cost / budgeting processes. This includes maintaining onsite inventory; developing computer systems support to the Plant; budgeting and tracking expenditures, in concert with appropriate line managers. The Training Manager is responsible to the RIM for providing the required training support to accomplish the scheduled hardware and
\
2-5
programmatic activities. This includes the identification of needs, in concert with appropriate line managers; development of training materials; presentation, evaluation and documentation of the training. The Licensing Manager is responsible to the RIM for providing the required licensing and engineering planning support for the scheduled hardware and programmatic activities. This includes the establishment of licensing strategies, in concert with the line managers; preparation of submittals in a timely manner, and coordination of meetings with NRC and other agencies. The Plant Modifications Manager is responsible to the RIM for the construction management of direct hire contractors at Rancho Seco. This includes managing a SMUD/ contracts non-manual organization to perform field engineering tasks in support of the direct hire labor contractors as well as monitoring and controlling the schedule and performance of the direct hire labor contractors. 2.2.2 Qualifications Deputy General Manager, Nuclear - John E. Hard Mr. Hard is an experienced chief executive-level, nuclear-experienced manager with 34 years proven performance in planning, directing, and analyzing complex operational and engineering projects and organizations. He has completed naval reactors training and has advanced degrees in nuclear physics. He is a recognized expert in the area of the utility industry regulatory processes. Mr. Hard has authored numerous papers presented at meetings of the AIF, ANS, ASCE, and PMI as well as being published in Electrical World and Public Utilities Fortnightly on the subject of nuclear regulation and utility management. He is a registered professional engineer in the fleids of Mechanical and Nuclear Engineering in the State of California. Restart Implementation Manager - Bill Bibb H. C. Bibb has over 30 years experience the nuclear power plant construction, test and start up, project management of plants during construction and general management and direction of power generating plants. He has held reactor operator and ! senior reactor operator licenses on large operating nuclear l power plants. He has completed numerous operations and technical training courses. He is a registered professional engineer (nuclear) in California. Plant Manager - B. G. Croley [ Mr. Croley has over 19 years experience in nuclear power generation in the areas of engineering, licensing, project management, training, and plant operations with Babcock and Hilcox, Westinghouse, and South Carolina Electric and Gas Company. He has been certified at the SR0 level by Westinghouse l 2-6 l
7, on the Zion station and has led efforts resulting in operating licenses for two nuclear facilities. (J ) In addition, he has managed the Nuclear Engineering Department and all V. C. Summer station functions while with South Carolina Electric and Gas. System Review and Test Program Olrector - Jim Field
- Mr. Field is the Nuclear Technical Support Superintendent at Rancho Seco. He has over 11 years experience in the Rancho Seco Technical Support Group. In addition, he has recently headed up the group which performed the Deterministic Failure Consequence Analysis described in Section 4.A.
Implementation Manager - J. R. Shetler Mr. Shetler has 15 years of Babcock and Wilcox PHR experience. This has included system design and procurement, startup/ test support, and outage maintenance and coordination activities. His outage coordination activities have spanned some ten outages over the last ten years including the role of outage manager.
- Quality Manager - S. R. Knight Mr. Knight has over 25 years experience in construction and operation of power plants with initial experience in design, project engineering and test engineering of U. S. Navy nuclear power plants. His recent experience has been in design, N licensing, construction, test and operation - of commercial
) generating plants including design reviews, safeguards, and v waste management, QA/QC, programs for inventory, material, and maintenance control.
- Executive Assistant - J. V. Vinquist Mr. Vinquist has 12 years of Nuclear Power Plant experience in increasingly responsible roles. He performed assignments as I&C Start-up Engineer, Assistant Electrical Maintenance Supervisor, Electrical Maintenance Supervisor, Maintenance Engineer, and Assistant Plant Manager - Technical support. He also obtained and maintained SRO license and periodically performed in capacity of Shift Supervisor. Prior to joining SMUD, he was a consultant assigned as Technical Staff Assistant to AGM, Nuclear at SMUD.
- Plant Modifications Manager - T. D. " Tom" Robertson Mr. Robertson has 20 years of experience in the construction field. While working for major design / construction firms, he had held various positions in project construction management involving nuclear power plants under construction, and modifications and maintenance to existing nuclear power plants.
Prior to his current assignment, he was the Unit 1 Construction Manager for the South Texas Nuclear Power Project. O 2-7 _ _~,m __.,--%., - . - _ . , - - - . , , , , ,p ,r
Training Manager - P. Turner Mr. Turner has worked within the training and nuclear fields for over 20 years. In addition to training assignments at the Tennessee Valley Authority, and the Institute of Nuclear Power Operations, Mr. Turner was Manager of the Nuclear Training Department at Kansas Gas and Electric company. Nuclear Engineering Manager - Greg Cranston Mr. Cranston has more than 20 years of proven management, supervisory, and mechanical / nuclear engineering experience in design, engineering, configuration control, scheduling, licensing and startup of nuclear power plants. Prior to joining SMUD, Mr. Cranston worked for a large architectural / engineering corporation where he was responsible for managing and supervising large, multi-discipline nuclear plant design engineering organizations. Mr. Cranston has co-authored ANS standards and is a member of the ASME and ANS. He has completed naval reactor training. He is a registered professional engineer in both Mechanical and Nuclear Engineering in the State of California. Licensing Manager - R. Ashley Mr. Ashley has 30 years of experience in atomic energy and nuclear power plant design and operation, with responsibilities in engineering licensing and project management. He has directed special licensing activities for two major nuclear plants and assisted on technical, scheduling, and licensing matters for several others. He has participated in developing and implementing the restart programs for two nuclear plants that received NRC shutdown orders. He is a registered professional Nuclear Engineer in the State of California. Support Services Manager - Roger E. Huebner Mr. Nuebner has 15 years of experience in the power generation field. While working for major engineering firms he had held various posttions in project management and control for projects involving nuclear power plants under construction, and modifications to existing nuclear plants. Prior to assuming his present role, he was Supervisor, Nuclear Cost Engineering at Rancho Seco. He is a registered professional engineer in the Commonwealth of Pennsylvania. O 2-8
i FIGURE 2-1 RESTART IMPLEMENTATION ORGANIZA110N c m3 > REY. 8 2/27/57 DEPUTY GENERAL MANAGER, NUCLEAR
- John Word i I I I I PUBLIC A '
MANAGEMENT EMERGENCY 3 R CE EXECUTIVE ASSISTANT PREPAREDNESS INFORMATION ASSISTANT MANAGER f + Brad Thomas 9
. Dick Weber + Don Mortin y + Kerry Sheorer + " "a v'"a"
D5"Em
- MAC + ( Vocont)
- PAC O INPO ,
- Nuclear Projects
- DECHTEL
^ CONTRACTOR RESTART
# MATRIXED FROM HQ IMPLEMENTATION MANAGER
+ 0111 Bibb I I I I T PRACTICES & ORGANIZATIONAL STAFF EXECUTIVE PROCEDURES EFFECTIVFNESS ASSISTANTS ASSISTANT COORDINATION
- Bob r,asby a George Roy SPECIAL PROJECTS a John Sepp
- A.J. ludury + Ron Lawrence + George Coward l I I I I I l PLANT PLANT tuPLEMENTATION TRAINING UCENSING NUCLEAR SUPPORT MAIJAGER MODinCATIONS MANAGER MANAGER MANAGER ENGINEERING SERVICES MANAGER MANAGER
+ Bob Croley - Tom Robertson + Jern Shetler + Pout Turner 7 Roy Ashley + Greg Gronston + Roger Huebner DEPUTY DEPUTY DEPUTY DEPUTY DEPUTY PLAtJT PLANT IMPLEMt:NTATION TRAINING UCENSING MANAGER MODIFICATIONS MANAGER MANAGER MANAGER
^ Bobby Day + (Vocont) - Lee f ossum
- Fred Gowers + (Vocont)
- Operations
- Modificottons + Area Coordinot en + Licensing
- Project Engineering . Materials
- Molntenonce & (Construction
- Planning
- Compliance
- System Design = Contracts Facilities Monogement) + Scheduling
- PRA Engineering + Budget
- Technical Services . Reporting + IAG
- RellobIlty Engineering
- Info Systems
- System Review & + EFIC Projects
- Purchasing Testing . TDI Projects
- Health Physics
1 1 1 2.2.3 Transition to the Target Nuclear Organization Our Restart and Implementation Organization is presently in effect as indicated in Section 2.2 and Figure 2-1. However, our transition from the Restart and Implementation Organization to the target or ; ultimate nuclear organization is now taking shape. The target nuclear organization draft is shown in Figure 2-2. Our objective is to fill all key positions with SMUD personnel prior to restart with one exception. Our acting Deputy General Manager (DGM), Nuclear, John Ward, is the only contractor who will continue to carry out his responsibilities through Restart. The SMUD Board has requested that Mr. Ward remain in the DGM Nuclear position at least long enough to assure that near-term (priority 2) and long-term (priority 3) items are scheduled and initiated and to assure continuity during our operational phase of 1987. An executive search for a permanent DGM Nuclear will commence in July 1987. At this time, qualif ted candidates have been identified for the Assistant General Manager (AGM), Nuclear Power Production and the AGM, Technical and Administrative Services and our plan is to appoint these two positions by March 31, 1987. All SMUD department managers and directors should also tentatively be in place by March 31, 1987. A SMUD individual hired for a key position will assume that position immediataly. If a contractor is presently filling the position at the time of the hiring, his duties will revert to that of a consultant to the SMUD individual until this process is no longer required. The implementation of this organization will be a major milestone in our progress towards restoring Rancho Seco to a routinely operating nuclear station. O 2-10
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4 [ '. LEVEL ONE ULTIMATE ORGANIZA110N FOR RANCHO SECO DGW NUCtIAR Pusue osaunw I, " " " * " - erUmano0N AW57mf 1 4 PWtegess, ase MassaWn
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-00R $ e F1re Protection e Pro t/ Procedures e Chemletry
- r. e Purchestng e Emergency M005 e AMl/1NPC/Etc. e ST e e Woote Management I interfees e System Eng.
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- Clarecel Poet -coctrkel e Doctried 1 de**8 per"mnsort (includes ALARA -1 4 C einet k Control
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&**',",,, e Orofthg nute and baits sums, o EQ engheerug prob-carenest e Seismic tems. (engineers) e ALARA Deelgi Review e Maintenance e Computer Operettone e Fuele Management
- Fionning e Date 8ese
< Adminletretion e Voice and Ooto Commun8cettene i e PC Camputer Services e Support of Softwara j Users mtv. g ----- M ATRIXED FRCW 1-10-87 HEAOCUARTERS JEW l FIGURE 2-2 2-11 - 1
2.3 ACTION PLAN ACTIVITY MANAGEMENT PROCESS The unique nature of the District's Action Plan combined with the time frame of implementation has necessitated the development and implementation of a process to accommodate adjustments to the plan while assuring the general objectives and direction of the program are maintained. To accomplish this, the District has established the necessary organizational structure, assigned appropriate authority and responsibility and developed the necessary guidelines to assure the program accomplishes its intended near-term and long-term objectives. The organizational elements instituted to meet these requirements include: 2.3.1 The establishment of the Recommendation Review and Resolution Board (RRRB) The Board is a nine member multi-disciplined group of individuals with nuclear experience and training drawn from SMUD, another utility with a B&W NSSS, the NSSS Vendor, and the Plant Architect Engineering firm. The functions of this Board are: (a) to screen recommendations for clarity and duplication, (b) evaluate issues and recommendations, and (c) recommend the appropriate disposition and oriority for the recommendation based on its technical merits. To guide the Recommendation, Review, and Resolution Board in making these technical assessments and prioritizing the implementation actions, the following guidelines have been established: Would implementation of the proposed recommendation: a) Reduce reactor trips b) Reduce challenges to safety systems c) Remain in nominal post-trip window d) Assure compliance with license requirements e) Minimize the need for operator action outside the control room within the first 10 minutes of an event f) Indicative of programmatic deficiency g) Significantly improve reliability / availability l If the recommendation meets any of these guidelines and is determined to be valid, the RRRB provides initial prioritization in accordance with engineering judgement as follows: O 2-12 l l
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. 1 CRITERIA FOR PRIORITIZING RESTART SCHEDULE
.1 ACTIONS TO BE COMPLETED PRIOR TO RESTART J OR COMPLETION OF THE RESTART TEST PROGRAM Assure plant remains in post-trip window Assure compliance with technical specifications Minimize the need for operator action outside the control room within the first 10 minutes of an event
.2 ACTIONS TO BE INITIATED AS PROMPTLY AS PRACTICA8LE, SCHEDULE DEVELOPED, RESOURCES ASSIGNED AND, MAINTAINED UNTIL COMPLETED (IT IS THE INTENT TO INITIATE THESE ACTIONS AS SOON AS. PRIORITY ONE ACTION COMPLETIONS MAKE RESOURCES AVAILABLE)
Enhance ability to remain in post-trip window automatically Reduce reactor trips Reduce challenges to safety systems Produce near-term programmatic benefits i .3 ACTIONS TO BE PROGRAMMED FOR THE LONGER TERM Improve reliability Improve availability Major programmatic enhancements In addition to- screening and validating recommendations, the RRRB inputs each recommendation to the master data base (QCI-12) which
- records and tracks each recommendation through its life cycle.
Once the RRRB has finished the validation process, the recommendation (valid or invalid) i s forwarded to different organizations based on its characteristics for action. The alternate paths are: (a) Programmatic recommendations are sent directly to the Performance Analysis Group (PAG) for disposition, and (b) system related recommendations are sent to the Systems Engineer, who in turn develops an integrated implementation and test plan for the system. This System Plan (System Status Report) i s then sent to the PAG for disposition. The PAG with the approval of the DGM, Nucl?ar, i determines the course of action for these recommendations Sn1 sends them to the appropriate departments for implementation through the Action Plan or the development of justification as to why the recommendation is invalid. , 2.3.2 The establishment of the Performance Analysis Group This group is made up of the Nuclear Department managers or their } designees that report directly to the Deputy General Manager, Nuclear. The function of this group is to review and determine the appropriate disposition, from a management perspective, of the recommended actions of the Recommendation, Review, and Resolution Board. . 2-13 i
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During this process, the recommendation is reviewed in light of existing program activities to determine whether the disposition of the recommendation (actions and priority) can be accommodated through existing programs or whether adjustments are necessary to assure the overall program objectives are met. This group also determines the department which will have the responsibility for implementing the necessary actions to satisfy the finding and recommendation. The Performance Analysis Group is also charged with monitoring the implementation of the Action Plan. The need for and approval of changes to the plan, which assure the near-term and long-term objectives are met, will be developed by the Executive Assistant and approved by the Performance Analysis Group and the Deputy General Manager, Nuclear. This includes changes to the priority of individual action items. 2.3.3 Impleme-tation and Close-out Approved actions are implemented by the appropriate line organization in accordance with the approved Quality Assurance manual and the existing approved department policies and procedures. The final step in the implementation process is the development of a closure package. This package contains sufficient information to describe the specific actions taken to implement the recommendation and where appropriate contains the actual implementation documentation. The closeout document is approved by the implementing department manager and forwarded to the Quality Department for final verification and closecut. Quality Assurance performs a verification of the implementation of the recommendation. This verification audit can apply sampling techniques where appropriate but will be of sufficient depth to assure the objectives of the recommendation have been met. When the determination is made that implementation is complete, the Quality Department documents this conclusion in the closure package and forwards it to the Performance Improvement Manager for logging and filing. 2.3.4 Action Plan Activity Tracking and Reporting The site Quality Department is responsible for the tracking of the Action Plan items. The tracking system employed by the site Quality Department has the necessary features to correlate these items or deficiencies it is being implemented to address. This is particularly important since each issue or deficiency may require several actions to accomplish closure, and a particular action may be required as part of the resolution of more than one issue. The status of each activity will be maintained, monitored, and reported on a weekly basis during program implementation prior to Restart and on a monthly basis following Restart Power Ascension testing. The Implementation Manager is responsible for maintaining the schedules for the implementation activities which impact plant hardware and management or programmatic issues. These schedules will ) l l 2-14 l i
t L i l I be updated frequently to satisfy the internal needs of plant outage : management but will be updated at least weekly to meet the external
@ interface requirements.
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2.4 REVIEW MEETINGS AND REPORTS The DGM, Nuclear will provide status reports to internal and external groups on a regular basis. Prior to Power operation, a special Restart Report will be prepared documenting the critical findings, and the implementations of the associated actions taken to address these findings during the implementation of the Action Plan. The periodic reports and meetings to communicate internally and externally are described below. 2.4.1 Internal On a monthly basis, the DGM, Nuclear will meet with the Rancho Seco Implementation Committee of the Board of Directors to review in detail, progress of the action plan. At the subsequent full Board meeting, he will present an overview of his report. On a monthly basis, the DGM, Nuclear will meet with the General Manager, Assistant General Managers and relevant staff to review in detail progress of the restart and performance improvement plan. Informally, the DGM, Nuclear will provide the General Manager with daily updates. On a monthly basis, the DGM, Nuclear will provide a summary status report for District employees. 2.4.2 External On 4 monthly basis, for the period beginning August, 1986, and extending to six months after the beginning of restart heatup, the DGH, Nuclear will meet with NRC staff and Region V to provide a formal status report on the progress of the Action Plan. On a monthly basis, for t4 period August,1986, and extending to six months after the beginninJ of restart heatup, the DGM, Nuclear will meet with the Independent Review Group to review in detail, progress of the Action Plan. Tht se meetings will continue on a quarterly basis for a one year period. On a monthly basis for the period August,1986, and extending to six months after the beginning of restart heatup, the DGM, Nuclear will send a monthly written report of Action Plan progress to the following organizations: American Nuclear Insurers Institute for Nuclear Power Operations B&W Owners Group Executive Committee Supervisors, Sacramento County Supervisors, Amador County Supervisors, San Joaquin County Chairman, California Energy Committee O 2-16
em 2.5 TRANSITION ACTIONS ( ) C/ The District's 1990 Plant Performance objectives are to achieve performance levels which will place Rancho Seco among the top performing nuclear plants in the United States. The District will modify the Nuclear Organization as required to achieve and maintain performance at this level by incorporating beneficial features of the systematic assessment program. Required changes to the organization will be in place prior to the disbanding of the temporary Restart and Implementation Organization (RIO). The termination of RIO is planned after sufficient actions have been taken on the long-term performance improvement actions to assure they 4 can be managed anti implemented by the normal line organization. Section 2.2 describes the current organization and our transition to the target nuclear organization. The long-term objectives will be achieved through the implementation of the performance improvement items identified in this document and the implementation of supplemental programs. In particular, the District intends to implement the programs necessary to: a) identify the components critical to power production. b) to monitor, trend, and evaluate unavailability contributors. c) to undertake a plant specific risk assessment. d) to implement those features of the precursor review program necessary to achieve a high quality lessons-learned program. The transition of the special NRC and IRG oversight activities for this Action Plan to those consistent with normal industry practices should occur when the restart actions are complete. A plan has been developed to assure completion of near-term and long-term actions and is described in Section 1.2.5. J 2-17
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i-s 3.0 PERFORMANCE IMPROVEMENTS UNDERNAY PRIOR TO THE DECEMBER 26, 1985 EVENT _ l
. In late 1984, the Board of Directors of the Sacramento Municipal
- Uttiity District recognized that the District had a significant L number of challenges facing them. Foremost among those challenges was the need for improvement in the operation of the Rancho Seco i Nuclear Generating Station. The Board recognized that the Rancho- l l Seco problems had developed over a number of years through the joint- r attitudes and performance of the Board, the Staff, and plant 4
personnel, t } To a large degree, these failures were made evident by the overburden !
; the District's staff felt in responding to the large number of I changes required to implement the TMI-2 lessons learned in areas >
4 including plant modifications, personnel performance, management i j systems, analytical capabillty, training, and organizational j structure. The Board also recognized that the dynamics of the pubitc power arena contributed significantly to the District's arrival at ! i its current situation. The Board was taking corrective actions on , these issues when a transient occurred on December 26, 1985, at } 'tancho Seco which emphasized the need for further action. 1. f During 1985, in recognition of the above situation, the Board ! embarked on an overall program to upgrade the District's organization and operations thereby improving the effectiveness and reliability of plant performance. The thrust of the improvement program was to deal : .I- with a large spectrum of management and organizational issues , j including. ; 1 l_ Estabitshment of a commitment to excellence in performance at ! i Rancho Seco, including strengthening the technical competency of i . the people and the organization. The effectiveness of interface activities within upper management and between departments. I
- Organizational streamlining, staff enhancement and other
{ organizational improvements. f Effective attention to detall. i l Upgrading of the training organization, training facilities, and ! training programs i i Estabitshment of a clearly defined maintenance program. i Establishment of an effective systematic troubleshooting program j 1
- Development of a comprehensive root cause analysis program !
4
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The intent of dealing with these issues was to elevate these areas of the operation to the level of excellence consistent with the charter of the Board and the expectations of the regulators, industry, and public. Detailed actions to implement the Board's desire for overall improvement were developed and each was assigned to a specific individual for completion on a specified schedule. Consistent with these plans a number of activities were initiated, beginning in the Spring of 1985, and the program continued through the summer. O O 3-2 __y _ .--i
3.1 PROJECTS UNDERWAY PRIOR TO DECEMBER 26, 1985
\
_,/ 3.1.1 Staffing and Organization Significant to the conditions determined in the 1984 study was the recognition of the sizable growth in plant staff, without a corresponding restructuring or expansion of the management staff. A new organizational structure was approved with six Nuclear Departments reporting to the DGM, Nuclear:
- 1. Nuclear Operations
- 2. Nuclear Engineering
- 3. Quality
- 4. Nuclear Training
- 5. Nuclear Licensing
- 6. Nuclear Projects The last three departments listed had previously been elements within the Nuclear Operations and Engineering Departments.
The Board approved this structure, and to provide the staff to fill
- new management and technical positions, nationwide recruiting efforts
- were mounted. All new key positions were staffed by early 1986.
Numerous other structural changes occurred within the various departments. The purpose and effect has been to reduce the diverse 4 managerial requirements upon Individual supervisors and superintendents by establishing new divisions and alignments which g bring similar functions under the direction of a single manager. Middle management is now better able to cope with the demands of the ' groups for whom they are responsible. He are seeing them spend more time on details while interacting with the personnel and projects coming under their purview. As a number of these people have been at Rancho Seco for less than a year, there has been a considerable
- Injection of new concepts and methods within the organization. This, coupled with the traditionally responsible and professional attitude of the Rancho Seco staff, has resulted in an overall attitude which j is receptive to the programmatic approach and committed to attention-to-detail and accountability.
3.1.2 Training Program ! 1. Management Restructure ! Previously, the Nuclear Training Department was an organization under the Operations Department Manager. It was recognized that this reporting level was inappropriate for the expanded importance of the training function and that it ought to be f elevated to departmental status to ensure top management involvement. i As of June 1985, the training organization became a department i answering directly to the Assistant General Manager, Nuclear. l i 3-3 I
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The position of Training Manager was established and filled from ; outside the District organization, bringing a new perspective ; and experience level to the department. These management and structural changes, together with appropriate training procedure and policy revisions, will result in management recognition of the training function as an integral part of plant operations and ensure effective coordination of the training function with all other nuclear I organization functions, i
- 2. INPO Accreditation Effort The District has been committed to INPO Accreditation for its training programs for the past several years in an effort to I improve the overall training program. The District is using a phased approach for this effort. The focus on obtaining accreditation for the various department functions is on a sequential basis, starting with operations.
The first phase, consisting of Senior Reactor Operator (SRO), Reactor Operator (RO), Shift Technical Advisor (STA), and Non Licensed Operator (NLO), received accreditation in April 1986. The remaining six training programs, which involve maintenance training, chemistry and radiological protection technician training, and technical support staff and managers training programs have been submitted for accreditation in June 1986. The accreditation process will ensure that the shortcomings identified in maintenance training are corrected. 3.1.3 Maintenance Program Prior to the December 26, 1985 event, the District had initiated a number of actions designed to improve its maintenance program. In addition, at the time of the event, several specific maintenance program enhancement actions were underway. These included: Search for an experienced individual from the industry to fill the Maintenance Manager position. Increased staffing levels authorized for maintenance in the 1986 Budget. This included allowances for Preventive Maintenance Supervisors and dedicated Preventive Maintenance Crews in the Mechanical, Electrical, and I&C Groups. Dedicated Planning Personnel authorized for the Electrical, Mechanical, and I&C Maintenance Groups.
- Formation of a centralized scheduling group to provide overall prioritization and integration of maintenance work with other plant activities.
A full-time consultant was reviewing and refining the Mechanical Preventive Maintenance Program. 3-4
3.1.4 Quality Assurance e p\ C/ Rancho Seco has undergone a number of changes in the qu'ality program and organization in the 1985 and early 1986 time period. These changes have been implemented in response to the independent analysis of an outside orga,nization (LRS). The two areas which have been impacted the most are Quality Control and Quality Engineering.
, Quality Control Inspectors from Nuclear Engineering Construction and Nuclear Operations were combined in mid 1985 and transferred to the
-; - Quality Organization. ,
; The Quality Control section consists of a Quality Control Supervisor with two QC Coordinators reporting to him. A total of 22 qualified inspectors that cover the area of I&C, electrical, mechanical, civil,
, NDE and concrete structures are now en board as District employees.
For outages, the inspegtors are augmented as needed by contract personnel. The QC inspectors are involved in source inspection, receipt inspection, construction and maintenance activities. f Quality Engineering was established as a new section in early 1985 and staffing to the current level of 10 professional engineers was
- , completed in January, 1986.,They were recruited from various A/E's
- and nuclear generating utilities. Quality Engineering is involved in the day-to-day maintenhnce operations, both corrective and preventive maintenance. The' design engineering review group-monitors the design control procedures and assures that quality requirements are added.to both design specifications and purchase requisitions. The addition
[ of Quality Engineering has increased the c'apability 6f the Q] organization and, therefore, its ability to perform /)tt intended function. j , 5 A major revision of the QA Manual was undertaken 1n3198S .> An extensive effort;was.nlade to upgrade the manual and separate it into
, two sections consisting of (1) 18-point policy sectyorf and (2) the i
Quality Assupane d roce'd uhs. The manual was r 61ehed and approved by the MSRC 2nd;NRC; Thepolicysectionwillbecoma7partofthe USAR. ,
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3.1.5 Systematic Troubleshooting g In March 1985, Rancho Seco was shutdown'for a scheduled 90 day ., refueling for Cycle 7. The plant was returning to service 94 days later when a RCS vent line cracked, causing a shutdown due to excessive lossrof prisy c'colant. The subsequent investigation and l repairs led to a greatly expanded IE Bulletin 79-14 pipe support program, which did nM allow restart until ] ate September. Between then and the end-of-the-year three reactor trips occurred. Upon each occurrence, a special sistamatic troubleshooting program (based upon the Davis-Besse NUREG 1154 Appendix B methods and criteria) was implemented. This was 1n addition to programs that were already in place such as Root Cause Analysis, the support work done by the B&W Owners Group Transient Analysis Program (TAP) team, and the Nuclear Operations Trip Report investigation. , l , ,/ 3-5 e , ty o s,- /^
. ...s,_ .__ _ _. _ . _ _ _ _ , _ , _ _
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l l l I I In each case, the implementation of the Systematic Troubleshooting i method led to a greater understanding of the event and determination I of corrective actions to preclude re-occurrence. This method was ; again instituted on December 26, 1985, as a first response to the ' December 26, 1985 overcooling event. 3.1.6 Root Cause Program Early in 1985, the Incident Analysis Group was established to provide independent analysis of events and activities to determine the programmatic root cause of each. They reported their findings to the Management Review Team which is made up of the Nuclear Department Managers and DGM, Nuclear. This program was quite successful during its first year in providing the independent, multi-disciplinary analysis necessary to produce useful root causes and programmatic determinations. 3.1.7 Activity Assessment It can be seen that basic problems had been recognized and initial actions taken before the December 26, 1985 incident. None of these programs had achieved the momentum to affect plant performance, though all have appropriately become key elements of this Action l Plan. Many actions associated with the District's initial program and the December 26, 1985 event analysis were completed prior to the development and submittal of this Action Plan. O O 3-6
p 4.0 RESTART AND PERFORMANCE IMPROVEMENT ACTION PLAN This section provides a description of the actions being taken to address the concerns raised by the December 26, 1985 event and the performance record of Rancho Seco. Section 4A provides a description of the systematic assessment-processes. Section 48 provides a description of the management, operations, and administrative process improvements for Rancho Seco functional areas as developed by the respective department managers. Section 4C prov* des a description of the plant modifications and maintenance improvements developed by the functional organizations. Section 4D provides a description of the Systems Review and Test Program. Note that a significant number of the actions committed to be completed prior to restart have been accomplished and are awaiting systematic closure by the QA process. O i 4-1
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4A SYSTEMATIC ASSESSMENT PROGRAM D This section provides an overview of the special tasks established to accomplish the systematic review of the physical plant, it's operating procedures, training, maintenance, and related areas impacting performance, including a look at industry and plant history. The details of these special tasks, and the procedural guidance for their implementation, are contained in QCI-12 " Plant Performance and Management Improvement Program (PP&MIP)". .The objective of these retrospective tasks is to assure that plant affecting deficiencies are identified and brought to management attention such that necessary corrective actions can be implemented. These special tasks include: Precursor Review Deterministic Failure Consequence Analysis
- B&W Owners Group - SPIP Program Plant Personnel Interviews 4A.a Commitment I The long-term benefits of having effective programs to accomplish I systematic assessments is clearly in the best intent of Rancho Seco.
l Prior to restart, these programs will be established as a part of the administrative process at Rancho Seco, to ensure that the benefits they provide will be applied to future issues, modifications, changes, and. conditions. O 4A-1
4A.1 PRECURSOR REVIEW PROGRAM 4A.l.1 Objective The objectives of the Precursor Review Program are to systematically review historical documents and recommendations for events or conditions and to determine their significance to Rancho Seco. From the events and conditions that are judged to be applicable and significant to Rancho Seco, a specific recommendation will be made to improve the affected plant area (design, operations, maintenance, etc.) to either preclude the occurrence or minimize the effect of the event or condition at Rancho Seco. The identified issues and improvement recommendations will be input to the Recommendations Review, and Resolution Board (RRR8) for disposition. 4A.l.2 Scope of Work The scope of work to be performed in the Precursor Review Program is divided into two parts.
- 1. Review of Past Trips and Transients on B&W-Designed Plants The review of transients on B&W-designed plants (transients are defined in the B&WOG SPIP Program) consists of the following:
- a. All Transient Assessment Program (TAP) Category C transients will be evaluated and investigated for their applicability and impact on Rancho Seco.
- b. All Category B TAP events will be reviewed to determine if any of the recommendations made are applicable to Rancho Seco and to determine whether, because of plant differences, the transient could have been more severe at Rancho Seco.
- c. All recommendations for Category A TAP transients will be reviewed to determine their applicability to Rancho Seco.
- d. All Rancho Seco transients, starting from the Rancho Seco
" light bulb" event in 1978, will be reviewed.
- e. Review NUREG-0667 " Transient Response of B&W Designed Reactors" for open recommendations.
- f. Review NUREG-0560 " Staff Report on the Generic Assessment of Feedwater Transients in Pressurizer Water Reactors Designed by B&W".
These events will be reviewed with recommendations or concerns identified and passed on to the RRRB. The review program described above will be completed before plant restart. O 4A-2 e
- 2. Other Document Reviews b In addition to the review of the TAP data, the following documents will be reviewed by a multi-discipline experienced team:
- a. Rancho Seco Licensee Event Reports and Occurrence Description Reports
- b. Significant Operating Experience Reports (SOER) issued by the Institute of Nuclear Power Operations
- c. Bulletins issued by the NRC Office of Inspection and Enforcement
- d. Notices /Ctrculars issued by the NRC Office of Inspection and Enforcement
- e. Babcock and H11cox Reports (Preliminary Safety Concerns, Site Instructions, and other relevant B&W reports)
This program provides for a reverse chronological review starting from 1985. Prior to start-up, documents dating back to March 1978 will bc reviewed. The Precursor Review team will access and make a recommendation as to the need and scope of any further reviews. 4A.I.3 Criteria and Methodology for Precursor Evaluation Each document will be reviewed to determine whether issues are applicable to Rancho i Seco. For each document a Precursor Review Checklist will be completed. For those issues which produce a recommendation, the Precursor Review Recommendation Form will be completed and forwarded to the RRRB. 4A.I.4 Schedule Evaluations of all recommendations will be completed and prioritized before plant restart. O 4A-3 i-
4A.2 DETERMINISTIC FAILURE CONSEQUENCE ANALYSIS 4A.2.1 Purpose The objective of the Deterministic Failure Consequence Analysis is to determine the consequences of failures of systems on power operation or post-trip response capability, and to evaluate related procedural guidance provided to the operators. The intent of the analysis is to identify areas where failures of plant systems or procedural inadequacies could potentially result in unnecessary reactor trips, unsatisfactory post-trip response, undue challenges to the operators, or challenges to the safety systems. Recommendations will be developed which improve plant reliability, post-trip response, and operator performance when or where inadequacies or enhancements are identified. 4A.2.2 Program Sccpe The effect of loss of electrical power, instrument air, and control power will be evaluated for impact on plant operations. These systems were chosen because failures in these systems closely approximate the consequences of most postulated plant system failures. The analysis will identify affected systems which challenge or adversely effect the capability to mitigate transient conditions. No attempt has been made to analyze every combination of failures which could occur, yet by starting with the assumed loss of an active component, and compounding the effect by assuming concurrent failure of components with common motive power or controls, a very wide range of likely partial loss of systems has been considered. While the methodology was applied, considering for the most part the individual loss of power, Instrument Air, or ICS/NNI, the results can be evaluated in light of the assumption that all three conditions exist simultaneously. Such a situation is the basis for the decision to install a class 1 Emergency Feedwater Initiation and Control (EFIC) system, plus class 1 air supplies to certain valves utilized in this event. This event would challenge the plant safety systems, but with these new features, it should not cause adverse transient conditions or control problems. 4A.2.3 Methodology Each system will be reviewed as described below:
.1 Loss of Electrical Power Teams will analyze each 480V bus, its source and loads. Each team will review electrical elementaries beginning at the end loads. Each breaker off the Motor Control Center (MCC) or panel will be " failed" inolvidually. (Note: The " failure" is an assumption, no physical positioning at circuit breakers, etc.,
will be attempted.) The consequence of failure of each load can then be determined. The process is repeated for all MCCs and 4A-4
l panels off a common 480V bus. Once the consequence of loss of (qi the individual loads is evaluated, the loss of the source (s) V will be analyzed. A similar analysis will be performed on the 120/125V buses, except that the failure will be assumed to include the inverter, battery, and alternate supplies. Upon completion of the analysis of the individual 480V buses, the loss of the 4160V bus and loss of the individual transformers to off-site power will be analyzed. Finally, the loss of off-site power will be analyzed. Electrical elementary drawings will be " yellow lined," identifying the breakers " opened" and affected components to ensure each load is addressed.
.2 Loss of Instrument Air An evaluation of the effects of loss of instrument air will be performed. Individual components (loads) on the Instrument Air
- System will be " failed" and the effect upon the plant determined. The entire system will then be " failed" to determine the effect on the plant. The P& ids will be " yellow lined" to ensure that each component and/or header is addressed.
.3 Loss of ICS/NNI U
The loss of ICS and NNI power supplies will be evaluated to determine failure states and resultant actions or suggested modifications necessary to establish a known safe state with little or no operator action. Appropriate drawings will be
" yellow lined" to ensure each component or parameter is addressed.
.4 IE Bulletin 79-27 Due to the number and scope of changes to ICS/NNI (the subject of this Bulletin) and the nature of the questions it asks (which are similar to the DFC approach), this Bulletin was evaluated as a part of the DFC scope, and utilized the DFC approach.
4A.2.4 Process Recommendations developed during the analyses will be submitted to the RRRB. Specific notes of those systems affected by the " failure" which lead to the recommendation will be made. These notes will be forwarded to the relevant system engineer. 4A.2.5 Schedule This project, including the evaluation of recommendations generated from the review, will be completed and prioritized before plant restart. ( 4A-5
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4A.3 B&W OWNERS GROUP PROGRAM - SAFETY AND PERFORMANCE IMPROVEMENT PROGRAM (SPIP) 4A.3.1 Background In January 1986, the B&WOG initiated a new concept in their management of B&WOG activities. Instead of having projects undertaken by individual committees, task forces, or working groups, which were directed only at the group's goals, it was decided that projects would be focused around a few major programs. These programs have since been identified to be: Trip Reduction and Transient Response Improvement Availability Improvement Regulatory Commitments Economic Benefits Each of these Programs has an organization to carry out its specific goals. Safety and Performance Improvement Program (SPIP) is the major element of the Trip Reduction and Transient Response Improvement Program. During the development of the Rancho Seco PP&MIP, reference was made to the B&WOG "STOP-TRIP" program. The "SPIP" and "STOP-TRIP" programs are the same, with SPIP being the programs finally adopted name. 4A.3.2 Objective The SPIP was established to improve B&WOG plant safety and performance. The program's stated objective is to: Reduce the number of trips and complex transients on B&WOG plants and ensure acceptable plant response during those trips and transients which do occur. The specific goals of the program are:
- 1. By the end of 1990, the average per plant trip frequency will be less than two per year.
- 2. By the end of 1990, the number of complex transients, as l classified by measurable parameters (Category "C") will be reduced to 0.1 per plant per year based on a moving three year average. (Note: SPIP Category "C" is similar to PP&MIP Priority "1" criteria).
The District is aggressively participating in the SPIP, as it does in the other B&WOG programs. Such involvement will ensure a broad perspective is taken with respect to plant improvements, as well as l allow other B&W owners to benefit from the Rancho Seco Plant Improvement Program. 4A-6
4A.3.3 Program Scope C The B&W Owners Group SPIP Program is similar in many ways to the District's Plant Performance Improvement Program. However, it is designed as an extension and expansion from previous B&W Owners Group (B&WOG) activities aimed at reducing the number, and severity of reactor trips which occur. Recently, the program scope has been expanded to collect information and develop the response to the NRC Assessment of the " Sensitivity" of the B&W Designed NSS. Thus, a number of programs are already underway and recommendations being prepared for evaluation and implementation by the member utilities. In addition, the B&WOG program is designed as an ongoing activity and thus will still be providing recommendations for plant improvement after the Rancho Seco Action Plan. The SPIP process includes the following major elements: Define concerns Prioritize concerns Integrate, schedule and perform projects Issue project reports with recommendations B&WOG Steering Committee approve and issue recommendations to owners Track implementation status with Recommendation Tracking System 4 l The most important aspect of the SPIP is the implementation of the recommendations at the B&WOG plants. b 4A.3.4 Methodology / Procedure To ensure effective SPIP implementation at Rancho Seco, the District has assigned a SPIP Coordinator (SPC) who provides the interface between the District and the B&WOG SPIP. Additionally, the District's coordinator is a member of the SPIP Management Team. This latter role is a unique opportunity to provide programs leadership and gain detailed insight that will aid implementation at Rancho Seco. The coordinator's role is shown schematically on Figure 4A-1 and is summarized as follows: Ensure SPIP recommendations are promptly input to the RRRB Follow up to ensure SPIP intent correctly interpreted Regularly report implementation status to B&WOG and to PAG
< Provide results of other PP&MIP actions to the B&WOG for consideration of generic applicability and integration with SPIP 4A.3.5 Schedule-Time of Performance In recognition of the fact that the SPIP will extend beyond Rancho Seco restart, any recommendations issued by the SPIP after startup will be addressed through the long-term program for similar treatment described in Section 4A.a. The District shall continue to fully support the objective, goals, and activities of the SPIP.
l 4A-7 l
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FIGURE 4A-1 SAFETY AND PERF0 NCE IMPROVE ENT PROGRAM (SP1P) SPC SUBMIT SPIP RECOMMENDATIONS TO RRR8 SUPPORT RRRB AND PAG REVIEW, MONITOR RESULTS REPORT DISTRICT STATUS KEEP PAG INFORMED REPORT STATUS OF NON-ON SPIP RECOMMENDATIONS OF DISTRICT STATUS SPIP DISTRICT RECOM-TO B&WOG VIA RECOMMEN- __ IN RTS MENDATIONS TO 8&WOG DATION TRACKING SYSTEM (RTS) O FOLLOWING RESTART SUB-MIT FUTURE SPIP RECOM-MENDATIONS TO COMMIT-MENT TRACKING SYSTEM O 4A-8
'4A.4 PLANT INTERVIEWS 4A.4.1 Purpose The interview program is to surface previously unresolved, but "known", problems which can (1) cause reactor trips and/or contribute to the severity of transients, and (2) degrade plant reliability or the optimal performance of the operating personnel. On this basis the result of the plant staff interviews were processed through the QCI-12 PP&MIP.
4A.4.2 Pro.iect Scope This program interviewed personnel from key plant and nuclear support functional groups. These persons were encouraged to identify systems, components, or operational problems and concerns of which they are aware and provide recommendations on how to resolve them. One hundred and fifty-seven volunteers were needed to satisfy the representative selection criteria based upon MIL-STD-1050. This requirement was met in all areas and a total of 180 volunteers have been interviewed. 4A.4.3 Interview Methodology The interview program covered a cross section of plant personnel and was intended to encourage personnel to identify issues or concerns of which they are aware that, when resolved, can contribute to optimal, reliable, or improved operation. Interstews were conducted utt11 zing O people who are skilled in the interview process. For the most part, these same people accomplished the Rancho Seco CRDR interview program, and they were assisted in preparing for the PP&MIP by consultants who are experts in industrial human relations. The details of this program are provided in Appendix I, which is extracted from QCI-12. An introduction and question form was prepared and presented to each interviewee prior to the interview. Each interviewee is asked for information about his/her background and is briefed as to the purpose of the interviews. The questions on the forms are then discussed one-by-one. Each interviewee is asked to expand on each answer until the interviewers feel no further meaningful information is available. The Interview Project Coordinator consolidates the compiled list of concerns / recommendations. He forwards the recommendations to the RRR8 using the Recommendation / Resolution sheets. The Interview Program Coordinator assures that the concerns and recommendations have been acted upon and dispositioned to the RRRB. 4A.4.4 Schedule The program interviews, including evaluation and disposttion of the recommendations will be completed prior to plant restart. Approximately 180 volunteers were interviewed. Some 1600 4A-9 l
recommendations were developed prior to consolidation to eliminate duplication. These recommendations have all been processed through the RRRS. O l l l O 4A-10 mr------ - - , ,- -_..-_ _--sm--,p,_ we--, -,g.,,-e--e,_ mm, , , , - -,m.e----- , -m m e- g- mr-ewm --,,--w~y - meg-
48 MANAGEMENT, OPERATIONS, AND ADMINISTRATIVE Pfl0 CESS IMPROVEMENT s This section identifies the actions developed by the District's department managers which are being implemented to address the deficiencies contributing to the performance record and the December 26, 1985 event. These actions form the framework for the implementation of the findings from the systematic review process. Each area contains specific commitments which address the deficiencies related to that area. This section addresses items scheduled to be accomplished prior to restart unless otherwise noted. 4 4 I I 48-1 l
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48.1 MANAGEMENT EFFECTIVENESS Management and management process effectiveness have a major impact on the ability to operate the Rancho Seco Nuclear Generating Station in a safe and reliable manner. The objectives of these actions are as follows:
- a. To develop guidelines and agreements by which the SMUD Board, as the governing entity, can improve its effectiveness in directing and monitoring the District's activities and obligations relating to the Rancho Seco Nuclear Generating Station.
- b. In light of significant reorganization and managerial changes, monitor the status of corporate headquarters management improvements and provide an assessment to the Deputy General Manager, Nuclear, General Manager and Board.
- c. To enhance the management process in support of the safe and reliable operation of the Rancho Seco Nuclear Generating Station.
48.1.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program 48.1.1.1 Review current executive level management practices and attitudes to ensure that executive level management processes support the safe and reliable operation of the Rancho Seco Nuclear Generating Station. The Management Process Review Group will obtain an opinion from the Independent Review Panel. Completion of this item closes QTE 25.0090 48.1.2 Actions that s;ay be Completed Prior to Restart (Near-Term Actions) 48.1.2.1.a Establish written performance measurement criteria and a performance review process for the General Manager (GM). Completion of this item closes QTS 25.0074 48.1.2.1.b Clarify the Board / General Manager working relationship in writing, including the reporting desired by the Board from the General Manager. Completion of this item closes QTS 25.0075 48,1.2.1.c Assess current corporate-support interfaces with the Nuclear Organization and make recommendations to the Assistant General Manager-Nuclear and the General Manager regarding improved management of interorganizational working relationships. Completion of this item closes QTS 25.0076 48-2
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I l l 48.1.3 Actions that may be Completed Prior to Restart (Near-Tern Actions) i N ! 48.1.3.a Develop and implement a Rancho Seco Business Plan for use by the Board of Directors. Completion of this item closes QTS 25.0077 48.1.3.b Establish a comprehensive, cohesive and clearly understandable set of GM and AGM-Nuclear policies and practices which provide upper tier direction for similar efforts at the functional manager and supervisory levels. Completion of this item closes QTS 25.0078 48.1.3.c Establish up-to-date functional organization charters and position descriptions which accurately reflect responsibilities authorities, and accountabilities for all organization functions and job classifications. Completion of this item closes QTS 25.0079 48.1.3.d - Upgrade management programs and practices in the areas of functional planning, decision making, problem solving and interdepartmental collaboration. i Completion of this item closes QTS 25.0080 48.1.3.e Establish appropriate management monitoring and control systems to ensure that all levels of department management are kept informed on important departoent performance trends or problem areas on a timely basis. At the same time, ensure that excessively burdensome administrative control systems are not perpetuated or introduced. Completion of this item closes QTS 25.0081 48.1.3.f Develop an employee communications program originating from the office of the AGM-Nuclear to ensure that all department employees are kept informed of District concerns, departmental priorities and performance progress on a timely basis and encouraged to feel that they are an important part of the Rancho Seco team. Completion of this item closes QTS 25.0082 48,1.3.g Develop a program for improving communications skills of Nuclear Department managers in presentations to the Board of Directors, the public and staff. Completion of this item closes QTS 25.0083 I l 4B-3 1-*yy- - e- e,- y--y--p--w,-,w.m--i-e.wwe=--rw--re* wp-,ww -------w+---em.-ww-+- -- .i------------v--g---,>wn .---- rea--w -- em
48.1.3.h Establish a department Human Resource Management program which includes:
- 1. identification of priority management development / training needs and the appropriate means for addressing each;
- 2. Identification of departmental priorities in terms of current vacancies and/or pipeline concerns;
- 3. engage more department management collaboration with the District's Human Resources organization in the recruitment / selection process.
Completion of this item closes QTS 25.0084 48.1.3.i improve Department media and community relations by establishing a more proactive media / community outreach program. Completion of this item closes QTS 25.0085 48.1. 3. J Improve Nuclear Department interfaces with all other Departments in the District by instituting additional interdepartmental communication and problem-solving processes on a regular basis. Completion of this item closes QTS 25.0086 48.1. 3. k Develop a Rancho Seco Facilities Master Plan. Completion of this item closes QTS 25.0087 i ( l O 48-4 i
A 48.2 QUALITY This section contains the list of outstanding action items which are scheduled for completion prior to restart. Their source is:
- 1. NRC Region V Open Items and Inspection Reports.
- 2. Amendment 1 of this Restart Plan.
- 3. Amendment 2 of this Restart Plan.
- 4. PAG approved items.
Action statements which reflect scheduled or completed work have been provided by the responsible manager or designee. Detailed closure requirements and completion dates or milestones for each item were simultaneously provided. 48.2.1 Action to be completed prior to Restart. 48.2.1.1 Reorganize the Quality function at Rancho Seco to enhance the Site Quality Assurance Department, providing increased focus in the following areas:
- a. Quality Enginsering
- b. Quality Control
- c. Surveillance
- d. Vendor Qualification and Source inspection
- e. Nuclear Program Audits CLOSURE REQUIREMENTS:
- 1. Organization and function charts reflecting reorganization.
- 2. Audit organization reporting to Site Quality Manager.
- 3. All Quality Department functions located at Rancho Seco.
- 4. Vendor program supervisor in place and two QE's assigned.
- 5. Quality Engineering Supervisor position filled by experienced consultant or SMUD employee.
Completion date: April 15, 1987 i i 48.2-1
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Completion of this item closes, or contributes to the closure of the following QTS items: 20.0112, 20.0127, 20.0351, 20.0393, 21.0050C, 21.0270, 26.0689 48.2.1.2 Develop and implement the procedures and processes necessary to independently verify the effective closure of the actions identified for the Action Plan. A Quality Tracking System (QTS) has been developed and is in operation to aid in accomplishing this task. CLOSURE REQUIREMENTS:
- 1. Issue QCI-12, Plant Performance and Management Improvenient Program.
Completion date: Complete Completion of this item closes, or contributes to the closure of the following QTS items: 20.0393, 21.0050C 48.2.1.3 Institute interim measures to strengthen the materials control at the Rancho Seco site. This action will provide additional assurance that materials being installed are properly documented and in compliance with the applicable codes and standards. CLOSURE REQUIREMENTS:
- 1. Issue joint directives (Site Quality Manager and Plant Manager) to establish interim controls for all Class 1 materials.
Completion date: Complete Completion of this item closes, or contributes to the closure of the following QTS items: 20.0049, 20.0393, 21.0001B, 21.0001C, 21.00010, 21.0050C, 26.0689 48.2.1.4 Institute interim measures to enhance the integration of QC planning with maintenance and construction instructions and activities. This action will assure effective and efficient quality inspection hold points are identified and implemented. Ct.0SURE REQUIREMENTS:
- 1. Quality Engineering personnel matrixed to planning organizations (corrective maintenance and modifications) to provide QC inspection planning for all Class I work at Rancho Seco.
48.2-2
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- 2. Revise AP.3, " Work Requests" to include inspection planning
/9 by Quality Department.
Completion date: Complete Completion of this item closes, or contributes to the closure of the following QTS items: 20.0351, 20.0393, 21.0001E, 26.0689 48.2.1.5 increase the Site QA Department staff to assure the added demands of the Action Plan and changes in responsibilities can be effectively implemented. CLOSURE REQUIREMENTS:
- 1. Add contractors to fill QE and EC contractor positions authorized in August, 1985.
Completion date: Complete Completion of this item closes, or contributes to the closure of the following QTS items: 15.0055-1, 20.0112, 20.0351, 21.0001E, 21.0001G, 21.0049, 21.0168, 26.0689 O N
. 48.2-3
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48.3 TRAINING This section contains the list of outstanding action items which are scheduled for completion prior to Restart. Their source is:
- 1. NRC Region V Open Items and Inspection Reports.
- 2. Amendment 1 of this Restart Plan.
- 3. Amendment 2 of this Restart Plan.
- 4. PAG approved items.
i Action statements which reflect scheduled or completed work have been provided by the responsible manager or designee. Detailed closure requirements and completion dates or milestones for each item were simultaneously provided. 48.3.1 Action to be Completed Prior to Restart 4a.3.1.1 Train operators to switch from AFW system operation to MFW system operation. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject O training.
Completion date: March 27, 1987 Completion of this item satisfies NRC Inspection Report 86-07 item 0-3 and closes the following QTS item: 25.0068. 48.3.1.2 Train operators on loss and partial loss of power to ICS and NNI including overall and specific plant equipment behavior during off-normal conditions in ICS, NNI and EFIC. CLOSURE REQUIREMENTS: i 1. Lesson plan and class rosters evidencing the subject training. Completion date: March 27, 1987 Completion of this item closes the listed QTS items and l J simultaneously satisfies NRC Inspection Report 86-07 Open Item j 0-9: j i 15.0001 Train operators on loss of ICS ' 15.0107 Train operators on use of multiple indications 15.0117 Train operators on 12/26/85 lessons learned 26.0035 Train operators on ICS failure 26.0050 Train operators on loss of power to NNI Signal conversion cabinet 48.3-1 l I
15.0118 Train operators using simulator on 12/26 event 15.0144 Revise training program to include 12/26 event 16.0009A Revise training program to include 12/26 event 18.0026 Train operators on loss of ICS/NNI 48.3.1.3 Train licensed operators in manual valve operation and MOV local operation. Subject training will include how the improper use of " cheaters" can lead to excessive torque / valve damage during manual operations. CLOSURE REQUlREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: Complete Completion of this item closes the listed QTS items and simultaneously satisfies NRC Inspection Report 86-07 Open Item 0-14: 15.0174 Train operators on local manual valve operation 15.01751 Train operators on use of " cheater" bars in valve operation 15.0176 Train operators on use of isolation valves 15.0032 Train operators on remote operator failure of FV-20527 & 20528 48.3.1.4 Train operators, during off-normal situations, to use correct and multiple indications when available. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: Complete Completion of this item closes the listed QTS items and simultaneously satisfies NRC Inspection Report 86-07 Open Item E-13: 15.0107 Train operators on multiple indications , 48.3.1.5 Train operators on pressurized thermal shock (PTS) concerns and l related pressure / temperature operating curves in Technical l Specifications. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: March 27, 1987 48.3-2
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Completion of this item closes the listed QTS items and O simultaneously satisfies NRC Inspection Report 86-07 Open Item 0-7: 15.0131A Train operators on RCS cooldown interpretation 16.0009A Train operators on RCS overcooling and PTS 48.3.1.6 Train operators on procedures and concepts which are designed to prevent overfill of the steam generator (OTSG) and introduction of water into the main steam lines. CLOSURE REQUlREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: March 27, 1987 Completion of this item satisfies NRC Inspection Report 86-07 Open Item 0-2. QTS Cross-reference: 25.0091 48.3.1.7 Train operators on all aspects of the post accident sampling system (PASS). CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: Complete Completion of this item will close itsted QTS items and j simultaneously satisfy NRC Deviation list 1/5/7 item 86-28-06: 26.0463 Train operators on pass operability 26.0050 Train operators on pass system 48.3.1.8 Train maintenance personnel in correct operation and maintenance procedures on the Essential HVAC (heating ventilating, air conditioning) system during normal and off-normal situations. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: March 27, 1987 Completion of this item will close listed QTS items and simultaneously satisfy NRC Inspection Follow-up list 3/66/103 Item 86-06-06. 26.0433 Train maintenance personnel on NSEB essential HVAC system 48.3-3
26.0021 Train all personnel on NSEB essential HVAC system 48.3.1.9 Revise administrative procedures to programmatically require training to revise lesson plans whenever tech specs are changed. CLOSURE REQUIREMENTS:
- 1. Revised AP indicating programmatic notification of Training upon Tech Specs revision which may require operator or other personnel training.
- 2. Internal Training Department document specifying programmatic lesson plan generation / modification upon tech specs change.
Completion date: March 27, 1987 Completion of this item satisfies NRC Follow-up List 7/66/103 item 83-12-05, and QTS 25.0065. 48.3.1.10 Maintain all Training records on suitable and reliable computer tracking system. CLOSURE REQUIREMENTS:
- 1. Production of a print-out evidencing recent training records. A signed memo from the Manager of Training which attests to the existence of a computer tracking system for training records will also suffice.
Completion date: Complete Completion of this item will satisfy NRC Follow-up list 24/66/103 Item 82-25-01, and QTS 25.0067. 48.3.1.11 Train operators and maintenance personnel on care and operation of station batteries. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: March 27, 1987 Completion of this item satisfles NRC Follow-up list 24/66/103 Item 86-07-15, and QTS 25.0069. 48.3.1.12 Train operators on the latest revisions to the Emergency Operating Procedures. CLOSURE REQUIREMENTS: O 48.3-4
(q V;
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: March 27, 1987 Completion of this item will close the listed QTS ltems and simultaneously satisfy NRC Inspection Report 86-07 Item 0-12. 15.0001 Train operators on Loss of ICS 15.0201 Train operators to inttlate investigation upon abnormal indications 15.0219 Train operators on EOPS 16.0012A Modify operator training on E0PS 26.0331 Train operators on revised EP's, OP's, CP's 15.0219 Train operators on revised EOPS 45.3.1.13 Train operators on specific issues and lessons learned from the 12/26/M overcooling event. CLOSURE REMIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training, or
- 2. Simulator training records evidencing subject training.
Completion date: March 27, 1987 Completion of this item will closed listed QTC ltems: 15.0117 Train operators on 12/26 event 15.0118 Train operators on 12/26 event using simulator 15.0144 Revise training program to include 12/26 event 16.0009A Train operators on RCS cooldown/ PTS 27.0045 Train operators on lessons learned from 12/26 event 45.3.1.14 Train operators on watch-standing principles including training on command and control concepts for supervisors, role and function of the STA. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training, or
- 2. Simulator training records evidencing subject training.
Completion date: Complete Completion of this item closes QTS ltems: 15.0149 Train operators on command and control concepts 48.3-5
l 15.0151A Train operators on control from CR priority over local 15.0151B Train operators on initiating investigation upon abnormal indication 15.0201 Train operators on initiating investigation upon abnormal indication 48.3.1.15 Train operators on Health Physics requirements associated with their job responsibilities, including aspects of entry into areas of unknown radiological conditions. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject j training.
Completion date: Complete Completion of tnis item will close QTS ltem: 15.00088 Train operators on entry into areas with unknown conditions 15.0202 Train operators on entry into areas with unknown conditions , 48.3.1.16 Train operators to maintain suction supply and leakage path for operating pumps. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: Complete Completion of this item closes the listed QTS ltems: 15.0059.1 Train operator to maintain supply and leakage path for operating pumps 48.3.1.17 Train operators on concept wheroin subcooling may be sufficient even though pressurizer level is abnormal. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: Complete O 40.3-6
Completion of this item closes the 11sted QTS items: 15.0064 Train operators on concept wherein subcooling may be sufficient even though pressurizer level is abnormal. 48.3.1.18 Train operators how to properly operate remote "BLPB-Seal in valves. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: March 27, 1987 Completion of this item closes the listed QTS ltem . 15.0115 Train operators to remotely operate BLP valves 48.3.1.19 Develop a program to incorporate Industry events into operator training programs. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class roster: evidencing the subject training.
Completion date: March 27, 1987 Completion of this item closes the listed QTS items: 15.01198 Develop a program to include LER's, SOER's, etc., into training programs. 44.3.1.20 Train operators on the use of the main feedwater (MFW) total flow recorders and the auxillary steam controller. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: Complete Completion of this item closes the listed QTS ltems: 15.129 Train operators on MFW total flow recorders and auxillary steam controller. 48.3.1.21 Train operators on concept of control from control Room vice local / control. p CLOSURE REQUIREMENTS: V 48.3-7
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: Complete Completion of this item closes the listed QTS Items: 15.0151A Train operators on concept of control from Control Room vice local / control , 15.0151B Train operators on concept of control from Control Room vice local / control 48.3.1.22 Train operators on use of C41 and A70 (Casualty Procedure 41 and Administrative Procedure 70). CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: Complete Completion of this item closes the listed QTS items: 15.0199.1 Train operators on use of C41 and A70 48.3.1.23 Train operators on 10/02/85 event CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: Complete Completion of this item closes the listed QTS items: 15.0331 Train operators on 10/02/85 event 48.3.1.24 Train operator on modified Emergency Plan. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: March 27, 1987 Completion of this item closes the listed QTS ttems: 15.0349 Train operators on modified emergency plan 48.3.1.25 Train operators on Instrument Air System (IAS) modification. O 48.3-8 i i I
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( f p I ," - j CLOSURE REQUlRDIENTS:
- 1. Lesson. plan and class rosters evidencing the subject training.
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Completion date: March 27, 1987 Completion of this item closes the listed QTS items:
;. ,f 19.00088 Train operators on IAS modifications
> 48.3.1.26 Train operators on modified Component Cooling Water (CCW) ptocedures.
h-1
; CLOSURE REQUIREMENTS:
- a. r, ..
- 1. Lesson plan and class rosters evidencing the subject training.
> Completion date: March 27, 1987 Completion of this item closes the listed QTS items:
26.0175 Train operators on modified CCH procedures n
48.3.1.27 Train operators on 120 VAC modifications
, CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: March 27, 1987 ,e e -c Completion of this item closes the listed QTS items: I-- 26.0267 Train operators on 120 VAC modifications L
< 48.3.1.28 Train operators on AFW-EFIC controls.
,f' CLOSURE REQUIREMENTS:
/
- l. Lesson plan and class rosters evidencing the subject
! training. Completion date: March 27, 1987 Completion of this item closes the listed QTS items: Y[ 26.0609 Train operators on AFH controls and valve controls ! Sti.3.1.29 Train operators on responsibiiitles associated with offluont discharges. 48.3-9
-+--,c, -+=ww--e,emm-e-..,w.
CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject training.
Completion date: March 27, 1987 Completion.of this item closes QTS 25.0072. 48.3.1.30 Train crafts on inspection of limitorque operators. CLOSURE REQUIREMENTS:
- 1. Lesson plan and class rosters evidencing the subject 1 training. i l
Completion date: March 27, 1987 Completion of this item closes the listed QTS items: 20.0132 Train craf ts on inspection of limitorque operators. , 1 O 1 O 48.3-10 l
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g 48.4 OPERATIONS AND OPERATING PROCEDURES This section contains the list of outstanding action items which are scheduled for completion prior to Restart. Their source is:
- 1. NRC Region V Open Items and Inspection Reports.
- 2. Amendment I of this Restart Plan.
- 3. Amendment 2 of this Restart Plan.
- 4. PAG approved items.
Action statements which reflect scheduled or completed work have been provided by the responsible manager or designee. Detailed closure requirements and completion dates or milestones for each item were simultaneously provided. 48.4.1 Actions to be completed prior to Restart. 48.4.1.1 Write Comprehensive procedure defining the policy for procedural compliance and procedural guidance.
- Cl.0SURE REQUIREMENTS:
- 1. Write review and approve procedure.
- 2. Implement procedure.
Completion date: March 31, 1987 Completion of this item satisfies open item 86-22-01 and 85-27-01 and simultaneously closes QTS 25.0066. 48.4.1.2 Correct procedural deficiencies identified during review of December 25, 1985 Transient and incorporate procedural changes resulting from System Engineers System Reviews. CLOSURE REQUIREMENTS:
- 1. Identify all effected procedures.
- 2. Modify / revise procedures.
Completion date: April 30, 1987 Completion of this item satisfies closes or contributes to the closure of the following QTS items: 22.0408 26.0022 26.0050 26.0079 O 48.4-1 e .y w--y.----wr- - --- w
26.0142 15.0006 15.0024 26.0288 15.0136A 15.01368 NCR Open Item IFI 86-07-07 48.4.1.3 Revise Administrative Procedures for Control of Emergency Procedures to assure verification and validation is performed. CLOSURE REQUIREMENTS:
- 1. Revise / implement AP48 & AP49.
Completion date: April 30, 1987 Completion of this item satisfies QTS 15.0140 and NRC Open Item 86-07-01. 48.4.1.4 Compare Emergency Operating Procedures to B&W Generic ATOG Technical Basis Document and revise E0P's to incorporate applicable information. CLO3URE REQUIREMENTS:
- 1. E0Ps are being revised in accordance with the New Technical Basis Document.
Completion date: Complete Completion of this item satisfies NRC Open Item RV-0-4, Inspection Report Section 0-7 and QTS 15.0139, QTS 15.0152B. 48.4.1.5 Revise operat39 procedures to reflect recommended topics of Regulatory Guide 1.33. CLOSURE REQUIREMENTS:
- 1. Revise / write 21 new procedures.
Completion Date: Complete Completion of this item satisfies: QTS 15.00448 QTS 26.0453 QTS 15.0097 QTS 15.0220 QTS 15.0229A QTS 26.0589 QTS 22.0389 QTS 15.0246A QTS 15.02468 QTS 18.00251 QTS 20.0349 QTS 22.0062 QTS 22.0064 QTS 22.0065 QTS 22.0066 QTS 22.0072 Qis 22.0400 QTS 23.0012 QTS 26.0348 QTS 26.0403 _ NRC Open Item 84-24-1-0 48.4-2 _- y
m' 48.4.1.6 Perform Valve Walkdown to verify configuration discrepancies between plant configuration and applicable documents and x procedures. CLOSURE REQUIREMENTS:
- 1. Review Walkdown findings and incorporate corrections to effected documents and procedures.
Completion Date: Prior to heat up. Completion of this item satisfies open items 86-13-02 and IN-85-66, QTS item 25.0070. 48.4.1.7 incorporate all procedure and training changes as a result of modifications implemented this outage. Ct.0SURE REQUIREMENTS:
- 1. Review modification packages.
- 2. Walkdown Modification.
- 3. Revise procedures, verify and validate E0P's
- 4. Simulator Exercises.
- 5. Actual plant operation.
Completion date: Prior to heat up. Completion of this item satisfies closes or contributes to the closure of the following QTS items: 26.0084 26.0085 26.0267 26.0329 26.0038 20.0166 19.0008A 48.4.1.8 Revise Fire Protection Administrative Procedures to ensure compliance with Appendix R requirements. Ct.0SURE REQUIREMENTS:
- 1. Rewrite / approve FPS procedure.
Completion Date: June 1, 1987 Completion of this item satisfies QTS 26.0587 0 48.4-3
- _ . m__-_. - , _,p9._,,. . , _ - _ _-_.--,_.p y
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3 48.5 MAINTENANCE PROGRAMS AND PROCEDURES
)
This section contains the list of outstanding action items which are scheduled for completion prior to Restart. Their source is:
- 1. NRC Region V Open Items and Inspection Reports.
- 2. Amendment 1 of this Restart Plan.
- 3. Amendment 2 of this Restart Plant.
- 4. PAG approved items.
Action statements which reflect scheduled or completed work have been provided by the responsible manager or designee. Detailed closure requirements and completion dates or milestones for each item were simultaneously provided. 48.5.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program. 48.5.1.1 Rework the Makeup Pump and return to service. CLOSURE REQUIREMENTS:
- 1. Repair the pump casing internals. Reconnect and align the q) pump, gear box and motor. Test the pump.
Completion date: Completed on schedule Completion of this item satisfies QTS 25.0088 4B . 5.1. 2 Perform refueling interval surveillance of snubbers. CLOSURE REQUIREMENTS:
- 1. Perform the described surveillance procedures.
Completion date: April 30, 1987 Completion of this item satisfies QTS 25.0059. 48.5.1.3 Perform biennial Diesel Generator inspection and replace turbo-charger. CLOSURE REQUIREMENTS: A) Perform "A" - Diesel Generator biennial inspection and turbo-charger replacement. B) Perform "A" - Diesel Generator biennial inspection and turbo-charger replacement. 48.5-1
.w.. , - - - - - - . , - - - .,y-w --w- .,-r-- , ey.-- - - - -, , ,-- r-.- e- -y -----% yw%,---.-- , -s y <q w - w- - --- --
I Completion date: A) is complete l B) April 30, 1987 Completion of this item satisfies QTS 25.0063 48.5.1.4 Define the critical items to be included in the Preventive Maintenance Program for: A) Manual Limitorque Operated Valves (105)
- 8) Manual Non-Limitorque Operated Valves (135)
C) Other Manual Valves important to process _ flow control in Class I and Steam Generator heat removal application (143). CLOSURE REQUIREMENTS:
- 1. Modify the PM Program AP-650 to include Manual Valve Preventive Maintenance Selection Criteria which identifies the critical manual valves.
Completion date: Complete Completion of this item will satisfy QTS 20.0132 48.5.1.5 Train crafts on inspection of Limitorque Valves. Complete Preventive Maintenance (PM's) on selected manual valves identified in 48.5.1.4. CLOSURE REQUIREMENTS:
- 1. PM all 570 manual valves as identified above.
Completion date: May 20, 1987 Completion of this item will satisfy QTS 25.0064 48.5.1.6 Complete the Rancho Seco Motor Operated Valve (MOV) refurbishment program NRC IEB #85-03. The following items will be addressed:
- a. Open/Close Torque Switch Bypass Limit Switches (IEN 86-29)
- b. Open/Close Limit Switches (IEN 86-71)
- c. Open/Close Torque Switches (IEN 84-10 and IEC 77-01)
- d. Switch Control for Butterfly and Three-Way Valves
- e. Overload Devices
- f. Interlocks with other Equipment (IEN 86-29)
- g. Position Indications (IEN 86-29)
- h. Operator Heaters (IEN 86-71 and lEB 79-01B)
- 1. Termination Splicing (IEN 86-53)
J. Environmental Qualification (EO) (IEN 86-03)
- k. Lubricant (IEN 79-03)
- 1. Pinion Gear Installation (IEN 885-22)
- m. Gearhead Worm Gear Orientation (IEN 83-02) 9 48.5-2
~\ l
- n. Stem Nut Lock Nut Staking (IEC 79-04)
- o. Hammering Effect from Non-Locking Gear Set (IEN 85-20)
\
CLOSURE REQUIREMENTS:
- 1. Refurbish and test all safety related MOV's. NRC IEB
#85-03. Completion of this item will satisfy QTS 19.0038. I Completion date: Prior to heat up.
48.5.1.7 a. Inventory Calibrated Test Equipment (CTE) and Calibrate and/or control use to prevent use of uncalibrated CTE.
- b. Write a procedtre to prevent use of uncalibrated CTE AP.33.
CLOSURE REQUIREMENTS:
- 1. Inventory as described above.
- 2. Procedure as described above.
Completion date: Complete ' Completion of this item satisfies QTS 25.0034 48.5.1.8 Provide a progran which assures current calibration of all in-plant instrumentation used in the performance of surveillance testing. CLOSURE REQUlREMENTS:
- 1. Memo of suitable proof of program which fulfills there requirements.
Completion date: Prior to heat up. Completion of this item satisfies QTS 25.0035 48.5.1.9 Complete rework of terminations in the Bailey Cabinets in the Control Room. ! CLOSURE REQUIREMENTS:
- 1. Evidence or memo stating completion.
Completion date: Restart Completion of this item satisfies QTS 25.0036 48.5.1.10 Perform all HVAC PMs. l CLOSURE REQUIREMENTS: 1 48.5-3
,,,n- ,- ,-r--i-e,r w- , - - - - 4 -v., ,- . - - - . , - - , - . ~ , . ----,r.
w% re.r-,,,,,.,--,em-.-,w- --- - -- - - r - -- ..-----=<---,=m v,ww,w-----. , -e
- 1. Memo stating completion.
Completion of this item will satisfy QTS Nos.: 26.0004. - Revise Maintenance Procedure for dampers and associated equipment. M-149 Damper Actuators - Complete 11/10/86 M-150 Damper Limit Switch - Complete 11/10/86 M-151 PM of Condensing Units - in review 11/5/86 26.0004. - Add doors HVAC units. 26.0023. - Function test each damper during refueling. 26.0025. - Maintenance procedure for Chlorine Gas Detectors. Completion date: Prior to heat up. , l 48.5.1.11 Complete the in-progess battery replacements (A,B,C,D,E,F) CLOSURE REQUIREMENTS:
- 1. Replace Banks A,B,C,D,E and F and startup test.
Bank B replacement during B DHS outage (currently 2-5-87 to 2-26-87) Deferred to 4-30-87. D.H. Outage. Address Generic Problem on post seals: Nuclear Engineering - Vendor on site 12-11-86 Resolution by 2-1-87 Complete awaiting ECN-1407 closecut. Completion date: Complete Completion of this item satisfies QTS 25.0050 48.5.1.12 Address NRC Insp Report 86-21-01 Resolution of A/B Diesel Light Bulb Failures (reference CCTS T860818006C) CLOSURE REQUIREMENTS:
- 1. Complete investigation and closure report. Closecut should occur by 1-1-87. Investigation completed 12/22/86.
Completion date: March 1, 1987 Completion of this item satisfies QTS 25.0051 9 48.5-4
x '48.5.1.13 Address NRC Insp report 85-08-01 Follow-up on Licensee Action i for HPI Breaker 4A04 (Reference CCTS T860428211C) CLOSURE REQUIREMENTS:
~
- 1. Complete investigation and closure report. Closecut should occur by 1-1-87 Completion date: March 1, 1987 Completion of this item satisfies QTS 25.0052 48.5.1.14 Address NRC Insp Report 85-16-01 NSCW Breaker Investigation Follow-up on Result and Corrective Action (Reference CCTS T860428212C)
CLOSURE REQUIREMENTS:
- 1. Complete investigation and closure report. Closeout should occur by 1-1-87 Completion date: March 1, 1987 Completicn of this item satisfies QTS 25.0053 48.5.1.15 Address 75tC Insp Report 86-07-12 Criteria and precedure for Placing Battery Banks on Equalize Charge.
CLOSURE REQUIREMENTS:
- 1. Revise procedures.
Completion date: March 1, 1987 Completion of this item satisfies QTS 25.0054. 48.5.1.16 Address NRC Insp Report 86-07-13 Non-conservative Practice of Correcting Spec Grav Readings. CLOSURE REQUIREMENTS: , 1. Revise procedures. Completion date: April 30, 1987 l Completion of this item satisfies QTS 25.0055 4B.5-5
48.5.1.17 Address NRC Insp Report 86-07-1A Practice of Equalizing Batteries Prior to Service Testing. CLOSURE REQUIREMENTS:
- 1. Write new procedures.
Completion date: March 31, 1987 Completion of this item satisfies QTS 25.0056 48.5.1.18 Address NRC Insp Report 86-08-05 Battery Surveillance Program (Reference CCTS - T861106112C) CLOSURE REQUIREMENTS:
- 1. Revise procedures.
Completion date: April 1, 1987 Completion of this item satisfies QTS 25.0057 48.5.1.19 Address NRC Insp Report 84-19-09 Upgrade of J Inverter Reliability CLOSURE REQUIREMENTS:
- 1. Eliminate J Inverter with ECN. Loads will be supplied from a more reliable source.
Completion date: April 1, 1987 Completion of this item satisfies QTS 25.0058 48.5.1.20 Address NRC Insp Report 86-07-16 Use of New Battery Banks Without Performing 8 H9ur Capacity Discharge Test. CLOSURE REQUIREMENTS:
- 1. Perform capacity discharge testing at the vendor's factory.
Closecut should be by 4-1-87 Completion date: April 1, 1987 Completion of this item satisfies QTS 25.0060 48.5.1 .21 CKT boards may have cracks under retainer clips. CLOSURE REQUIREMENTS:
- 1. Conduct inspections and repairs.
Completion date: Prior to heat up. O 48.5-6
Completion of this item satisfies QTS 25.0084 O 48.5.1.22 Address NRC Open item 8403X0 Cracks in Switch Jaws of Westinghouse MG-6 Relays (Reference CCTS T861106108C). l CLOSURE REQUlREMENTS:
- 1. Complete investigation of defective relays. Paperwork closure package ready by 3-1-87 Completion date: March 1, 1987 Completion of this item satisfies QTS 25.0061 48.5.1.23 Address NRC Open item 8520P Defective UVD on General Electric type AK breakers.
CLOSURE REQUIREMENTS: Closecut should be by 3-1-87 Completion date: March 1, 1987 Completion of this item satisfies QTS 25.0062 O l l l l l O 48.5-7 e - - -w-w ww-w-www-wewwr-m,,- , - - - - - -----m-w -e*-geww g---f -.-r--------wwmey,-p yo-+w9-w - r,- wr-w w wr-e w ww-meme-m
p 48.6 HEALTH PHYSICS AND RADIOLOGICAL CONTROLS This section contains the list of outstanding action items which are scheduled for completion prior to Pestart. Their source is:
- 1. NRC Region V Open Items and Inspection Reports.
- 2. Amendment 1 of this Restart Plan.
- 3. Amendment 2 of this Restart Plan.
- 4. PAG approved items.
Action statements which reflect scheduled or completed work have been provided by the responsible manager or designee. Detailed closure requirements and completion dates or milestones for each item were simultaneously provided. 48.6.1 Action to be completed prior to Restart. 48.6.1.1 Relieve operators of special HP duties. CLOSURE REQUIREMENTS:
- 1. Assign a " Coverage" RP Tech on shift to relieve operators of Special HP duties.
Completion date: Complete QTS cross reference: 25.0044 48.6.1.2 Prepare a procedure for Health Physics Technicians to use for entry into unknown radiological conditions. CLOSURE REQUlREMENTS:
- 1. Issue Administrative Procedure AP 313-5, Rev. O, " Unknown Radiological Conditions Guidelines."
Completion date: Complete QTS cross reference: QTS 15.0008A 48.6.1.3 Revise setpoints for plant gaseous effluent monitors to ensure unambiguous indicators. CLOSURE REQUIREMENTS:
- 1. Issue AP 165, Rev. 24 to correct monitor discrepancies.
Completion date: Complete 48.6-1
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, - - .,m. .-, -_ .,, ,y .,%yp,__,-.,, .-,,-w. - ., ,. __. ,,.-, _ ,_ .. .,c-,-,-.m,.
QTS cross reference: QTS 15.0011 48.6.1.4 issue a Radiological Event Directions Manual to provide more guidance for abnormal situations. CLOSURE REQUIREMENTS:
- 1. Issue AP 313, the " Radiological Event Directions Manual."
Completion date: Complete QTS cross reference: QTS 15.0318 48.6.1.5 issue new manuals to separate event and instrument procedures i from the Radiation Control Manual. CLOSURE REQUIREMENTS:
- 1. Issue AP 313 " Radiological Event Directions Manual."
Completion date: Complete
- 2. Issue AP 311, " Radiation Detection Instruments."
Completion date: Complete QTS cross reference: 25.0038 48.6.1.6 Organize and maintain the Health Physics Support and Environmental Programs group. CLOSURE REQUIREMENTS:
- 1. Organization is fully staffed or positions are filled by qualified contractor personnel.
Organization Completion Date: Complete Staffing Completion Date: December 31, 1987 Completion satisfies, in whole or in part: NRC item 86-20-01 and enforcement conference (June 20, 1986) commitment. QTS cross reference: 25.0045, 25.0037 48.6.1.7 issue revised ALARA Manual. CLOSURE REQUIREMENTS:
- 1. Completed document approved by the Deputy General Manager, Nuclear and the General Manager.
O 48.6-2 - r -----rm - - - -- ---- - - - - . - -. --
s 2. Completion of appropriate training.
\
Completion date: March 20, 1987 Completion satisfies, in whole or in part: NRC item 86-20-01. QTS cross reference: 25.0039, 25.0031 48.6.1.8 Develop technical support and auditing management processes and procedures to interface HPSEP and Radiation Protection groups. CLOSURE REQUIREMENTS:
- 1. Approval and issuance of the ALARA Manual.
- 2. Approval of management processes by the Plant Manager.
- 3. Approval of procedures by the HPSEP and Radiation Protection Superintendents.
Completion date: May 29, 1987 Completion satisfies, in whole or in part: NRC item 86-20-01. QTS cross reference: 25.0043 48.6.1.9 Respond to NRC Violation Enforcement 86-06-05: Evaluate actions with implementation of AP.305 and H2 PSA-7. CLOSURE REQUlREMENTS:
- 1. Issue Rev. O of Annunciator Procedure H2PSA Completion date: Complete QTS Cross
Reference:
25.0026 O 48.6-3
48.7 10CFR50 APPENDIX l DISCHARGE GUIDELINES V This section contains the list of outstanding action items which are scheduled for completion prior to Restart. Their source is:
- 1. NRC Region V Open Items and Inspection Reports.
- 2. Amendment I of this Restart Plan.
- 3. Amendment 2 of this Restart Plan.
- 4. PAG approved items.
Action statements which reflect scheduled or completed work have 4 been provided by the responsible manager or designee. Detailed closure requirements and completion dates or milestones for each item were simultaneously provided. 48.7.1 Action to be Completed Prior to Restart 48.7.1.1 lasue Complete Revisions of the Off-Site Dose Calculation Manual (0DCM) CLOSURE REQUIREMENTS:
- 1. Completed document with all supporting documentation approved by Management Safety Review Committee (MSRC).
Draft for Review Completion Date: February 6, 1987 Approval Completion Date: April 3, 1987 Completion satisfies, in whole or in part: NRC items 85-03-03, 86-15-01, and 86-15-09; ANI/MAELU 86-1, item I. QTS cross-reference: 25.0023 48.7.1.2 issue the Off-Site Dose Calculation Implementing Procedures Manual. ! CLOSURE REQUIREMENTS: f 1. Completed document approved by the Plant Review Committee (PRC).
- 2. Completion of appropriate training.
Draft for Review Completion Date: April 10, 1987 Approval Completion Date: May 22, 1987 lO 48.7-1 yr - , ., y_ _,, y.-.. - . _ _ - - - . - -- -y,,-y. ,.%-._, _ _ - . . -.~ , , , ,y
I Completion satisfies, in whole or in part: NRC items 86-15-01, 86-15-09, and 85-03-03; ANI/MAELU 86-1, item III. QTS cross-reference: 25.0027 48.7.1.3 Issue the ODCM Database Manual CLOSURE REQUIREMENTS:
- 1. Completed document and all supporting documentation reviewed by the Management Safety Review Committee (MSRC).
Draft for Review Completion Date: April 10, 1987 Approval Completion Date: May 22, 1987 Completion satisfied, in whole or in part: NRC item 86-15-13 and ANI/MAELU 86-1, item I. QTS cross-reference: 25.0028 48.7.1.4 Complete the 1Iquid offluent and 50-mile radius Iand use census and pathway evaluation. CLOSURC REQUIREMENTS:
- 1. District Review of services vendor draft reports.
- 2. Issue final land use census reports for 1986 and USAR update.
- a. Five-mile radius - gas.
- b. Five-mile radius - liquid.
- c. USAR 50-mile radius.
Completion Date: June 19, 1987 - Completion satisfies, in whole or in part: NRC item 85-03-03 and ANI/MAELU 1tems I and III. QTS cross-reference: 25.0029 48.7.1.5 Revise the 00CM Database Manual to include results of the liquid affluent and 50-mile radius land use census and pathway evaluation. CLOSURE REQUIREMENTS:
- 1. Completed document reviewed by the Management Safety Review Committee (MSRC).
Completion Date: June 26, 1987 Completion satisfies, in whole or in part: NRC items 85-03-03 and ANI/MAELU 86-1, items I and III. 48.7-2
QTS cross-reference: 25.0030 b' 48.7.1.6 issue Off-Site Exposure Control Management Process and Manual. CLOSURE REQUIREMENTS:
- 1. Completion and issuance of the revised ALARA Manual.
- 2. Completed document approved by the Deputy General Manager -
Nuclear and the General Manager.
- 3. Completion of appropriate training.
Draft for Review Completion Date: March 27, 1987 Approval Completion Date: May 22, 1987 Completion satisfies, in whole or in part: NRC enforcement conference (June 20, 1986) commitment. QTS cross-reference: 25.0032 43.7.1.7 Complete Off-Site Dose Calculation Sensitivity analysis. CLOSURE REQUIREMENTS: Analysis complete QTS cross-reference: 25.0033 Completion Date: Prior to heat up. 48.7.1.8 Identify Systems Affected by Health Physics and Radiation Protection Software. CLOSURE REQUlREMENTS:
- 1. Issue document which details the affected systems and l software. i
- I Completion Date: April 17, 1987 Completion satisfies, in whole or in part: ANI Bulletin 86-1, item IV.
QTS cross-reference: 25.0046, 25.0040 48.7.1.9 Determine Quality Classifications and Establish Performance Standards for Health Physics and Radiation Protection Software. CLOSURE REQUIREMENTS:
- 1. Issuance of Off-site Exposure Control Management Process and Manual.
- 2. Issue documented classifications and standards.
O 48.7-3 l
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, , , , , , - , ---.y yy.-- - - , ym . -, s.-,-- -.--. ,--, , v.--,- -
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Completion Date: May 22, 1987 Completion satisfies, in whole or in part: ANI Bulletin 86-1, item II and IV. QTS cross-reference: 25.0047, 25.0041 48.7.1.10 issue Procedures for Control of Health Physics and Radiation Protection Software Development, Modification, Access, Benchmarking. CLOSURE REQUIREMENTS:
- 1. Completed document approved by the HPSEP and Radiation Protection superintendents.
Completion Date: June 12, 1987 Completion satisfies, in whole or in part: ANI Bulletin 66-1, item II. QTS cross-reference: 25.0048, 25.0042 48.7.1.11 Prepare Procedures for Calculation of Off-Site Dose Associated with Unmonitored and Unplanned Releases (NRC RG 1.21) of radioactive materials. CLOSURE REQUIREMENTS:
- 1. Procedures approved.
Completion Date: Prior to heat up. QTS cross-reference: 25.0049 48.7.1.12 Revise Technical Specifications Associated with 1CFRSO Appendix 1 Compilance. CLOSURE REQUIREMENTS:
- 1. Delivery of completed revisions to Regulatory Compliance.
l Completion Date: Complete l Completion satisfies, in whole or in part: RWC 86-696; NRC item 86-15-11. QTS cross-reference: 25.0017 48.7.1.13 include Additional Sampling / Monitoring Points for the Liquid Effluent Pathway in the Radiological Environmental Monitoring l Program. CLOSURE REQUIREMENTS: 48.7-4
p* 1. Sampling / monitoring points are included in the radiological environmental monitoring program routine collection and (s analysis program. Completion Date: Complete Completion satisfies, in whole or in part: RHC 86-696. QTS cross-reference: 25.0018 48.7.1.14 Perform Updated Error Analysis on Systems Associated with Radioactive Effluent Release and Monitoring. CLOSURE REQUIREMENTS:
- 1. Include analysis in Semiannual Radioactive Effluent Report, January-June 1986.
Completion Date: Complete Completion satisfies, in whole or in part: RWC 86-696. QTS cross-reference: 25.0019 48.7.1.15 Revise the Updated Safety Analysis Report to Reflect Actual Liquid Effluent Control Practices. CLOSURE REQUlREMENTS:
- 1. Delivery of completed revisions to Regulatory Compliance.
Completion Date: April 2, 1987 Completion satisfies, in whole or in part: NRC item 86-15-11; RWC 86-696. QTS cross-reference: 25.0020 48.7.1.16 Develop and implement the changes in Radiochemistry methods and controls necessary to satisfy the listed NRC open items. CLOSURE REQUIREMENTS:
- 1. Issue or revise the following documents:
, AP.306 V-13, Rev. 3 AP.306 III-4, Rev. 13 SAR 828 l SAR 844 ! AP.305-13, Rev. 22 AP.310-lC, Rev. 4 I l 48.7-5
AP.310-2C, Rev. 2 AP.310-2CA, Rev. O AP.310-3C, Rev. 7 Completion date: Complete Completion of this item satisfies Enforcement 86-15-01, Enforcement 86-15-09, Enforcement 86-15-02, QTS Cross j
Reference:
25.0021. 48.7.1.17 Review and revise the off-site discharge calculations manual. CLOSURE REQUIREMENTS:
- 1. Issue AP.310-lL, Rev. 4. Completed October 8, 1986
- 2. Issue AP.310-2L, Rev. 4. Completed October 8, 1986
- 3. Issue AP.310-2LA, Rev. O. Completed October 8, 1986 l 4. Issue AP.310-3L, Rev. 7 QTS Cross
Reference:
25.0022 Completion Date: Prior to heat up. O O 48.7-6
48.8 EMERGENCY PREPAREDNESS This section contains the list of outstanding action items which are scheduled for completion prior to Restart. Their source is:
- 1. NRC Region V Open Items and Inspection Reports.
- 2. Amendment 1 of this Restart Plan.
- 3. Amendment 2 of this Restart Plan.
- 4. PAG approved items.
Action statements which reflect scheduled or completed work have been provided by the responsible manager or designee. Detailed closure requirements and completion dates or milestones for each item were simultaneously provided. 48.8.1 Action to be completed prior to restart 48.8.1.1 Modify the Emergency Plan (and appropriate procedures) to include the on-shift HP (RP) technician as part of the Emergency Team. CLOSURE REQUlREMENTS::
- 1. AP 514 to be revised to include RP tech in Emergency Team.
This item closed 12/03/86. See AP 514 Section 2.2.
- 2. Revise AP 500 (Plan) at 5.4.19 to include RP tech as part of Emergency Team.
Completion date: Complete Completion of this item will simultaneously close QTS No. 15.0025 48.8.1.2 Revise / update county and state voice notification system. CLOSURE REQUIREMENTS:
- 1. Develop plan to upgrade system.
Completed.
- 2. Implement plan to upgrade system.
Completion date: June 30, 1987 Completion of this item will simultaneously close QTS No. 15.0025E and 26.0624 x i 48.8-1
~
1 I
48.8.1.3 Train all operators on Emergency Plan modifications. CLOSURE REQUIREMENTS:
- 1. This action interpreted as a continuing need, therefore and EPIP revision to implement changes in training as a result of plan or procedure changes.
Completion date: May 10, 1987 Completion of this item will close QTS 15.0349 l i . O 1 0 48.8-2
48.9 HUMAN FACTORS This section contains the list of outstanding action items which are scheduled for completion prior to Restart. Their source is:
- 1. NRC Region V Open Items and Inspection Reports.
- 2. Amendment 1 of this Restart Plan.
- 3. Amendment 2 of this Restart Plan.
- 4. PAG approved items.
Action statements which reflect scheduled or completed work have been provided by the responsible manager or designee. Detailed closure requirements and completion dates or milestones for each item were simultaneously provided. 48.9.1 Changes to the Main Control Room invalidate the CRDR. 48.9.1.1 Develop a Human Factors Program and apply the program to all MCR modifications before restart. CLOSURE REQUIREMENTS:
- 1. Modify Nuclear Engineering Procedures to incorporate a Human O. Factors Program.
Completion date: Prior to Control Room modification implementation. Completion of this item satisfies QTS 26.0527 l l
) )
4B.9-1
r 48.10 MANAGEMENT INFORMATION SYSTEM f-This section contains the list of outstanding action items which are scheduled for completion prior to Restart. Their source is: 1 NRC Region V Open Items and Inspection Reports.
- 2. Amendment 1 of this Restart Plan.
3 Amendment 2 of this Restart Plan.
- 4. PAG approved items.
Action statements which reflect scheduled or completed work have i been provided by the responsible manager or designee. Detailed closure requirements and completion dates or milestones for each item were simultaneously provided. 48.10.1 Action to be completed prior to Restart. 48.10.1.1 Provide information management services support for restart activities. This support includes, but is not limited to, the following activities: Development of the Quality Tracking System and Master Tracking System Upgrade data communications
- Upgrade mainframe computer and storage Development of the Material Management System Standardization of site hardware / software to allow for future compatibility Installation and development of the Training information System
- Enhancement and support for work management personal computer based applications
- Installation of local area network to support Quality Tracking System and workflow
- Continue and improve support of existing applications Ct.0SURE REQUIREMENTS:
- 1. Support and service activities will continue during the restart program.
QTS cross-reference: 25.0006 ! 48,10.1.2 Develop plan for implementation of Generic letter 84-28 commituonts. Ct.0SURE REQUIREMENTS: . 48.10-1
- l. Plan fulfilling above requirement.
Completion date: Prior to heat up. QTS cross-reference: 25.0007 O' O 48.10-2
~
l
48.11 C00911TMENT MANAGEMENT Commitment Management is an important aspect of the District's interface with "outside" agencies as well as for the management of the District's day-to-day activities. An effective commitment management program results in improved management controls and effectiveness, and a greater degree of confidence that regulatory requirements are met and remain implemented. The development of an effective commitment management program is an iterative and long term process. This process involves management support, programatic controls, personnel training and software system development. In order to ensure optimum effectiveness, a long-term perspective shall be taken with regard to commitment management; the immediate actions specified herein are designed to support this perspectiva. 48.11.1 Action to be Completed Prior to Restart 48.11.1.1 Revise commitment management procedural controls to improve the tracking function and address ongoing compliance. CLOSURE REQUIREMENTS:
- 1. Issue Directive on control of commitments to regulatory agencies.
Completion date: Complete
- 2. Issue Directive on commitment management process.
Completion date: Complete
- 3. Issue Management process on regulatory commitment management describing continued compliance.
Completion date: Prior to release to Olspatcher. j QTS reference: 25.0008 48.11.1.2 Develop a commitment management system software design to support the District's long-term goals. I CLOSURE REQUIREMENTS:
- 1. Issue a new commitment management system functional specification.
Completion date: March 15, 1987 QTS reference: 25.0009 48.11-1
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i i 48.11.1.3 Develop a long-term plan for commitment management. CLOSURE REQUIREMENTS:
- 1. Issue Commitment Management Program Plan.
Completion date: March 15, 1987 QTS reference: 25.0010 - 48.11.1.4 Verify the commitment tracking system action item database with respect to current known commitments. CLOSURE REQUIREMENTS:
- 1. Documented review of data transfer from original tracking system to current system.
Completion date: May 1, 1987
- 2. Documented review of status of commitments on the current system.
Completion date: May 1, 1987 ,
- 3. Document commitment status with respect to k.iown NRC ,
outstanding items. Completion date: May 1, 1987 QTS reference: 25.0011 , 48.11.1.5 Verify all commitments required prior to Restart are complete. , COMPLETION REQUIREMENTS:
- 1. Generate QTS documentation identifying commitment completion.
Completion date: Prior to release to Dispatcher. QTS reference: 25.0012 l 1 0 48.11-2
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4> l J k j 48.12 CONFIGURATION MANAGEMENT
\ ,
- This section contains the list of outstanding action items which n are scheduled for completion prior to Restart.
- c. ,
4 Their source is: c -
- l. NRC Region V Open Items and Inspection Reports.
f~ '
- 2. Amendment 1 of this Restart Plan.
- 3. Amendment 2 of this Restart Plan.
//. 7
- 4. PAG approved items.
]
l ' ,I'." 1 : Action statements which reflect scheduled or completed work have
^;-
' been provided by the responsible manager or designee. Detailed closure requirements and completion dates or milestones for each A ,< item were simultaneously provided.
48.12.1 Actions to be completed prior to Restart. 4 - 45.12.1.1 Verify that control room drawings are current in accordance with existing procedures. CLOSURE REQUIREMENTS: i l. Nemorandum stating that verification is current. G , Completion date: Prior to release to Dispatcher. QTS reference: 25.0013 48.12.1.2 Make the necessary modifications to the design change process to ensure that design change.- are incorporated into all operating procedures in a timely manner (commitment 48.4.1.6 in Amendment 1). CLOSURE REQUlREMENTS:
- 1. Revised procedure issued.
Completion date: Complete QTS reference: 25.0014 i 'j' 48.12.1.3 Establish a program to verify the identification consistency of as-installed components Class I, P&lDs and the Master Equipment i - 4 ! ' List, and to initiate corrective action for inconsistencies which are identified. CLOSURE REQUIREMENTS: i l r > 48.12-1
l
- 1. Program Plan and Implementing Procedure issued. i Completion date: April 1, 1987 !
l QTS reference: 25.0015 l l l O O 48.12-2
r 48.13 MATERIALS MANAGEMENT k This section contains the list of outstanding action items which ; are scheduled for completion prior to Restart. ' Their source is:
- 1. NRC Region V Open Items and Inspection Reports.
- 2. Amendment 1 of this Restart Plan.
- 3. Amendment 2 of this Restart Plan.
- 4. PAG approved items.
Action statements which reflect scheduled or completed work have i been provided by the responsible manager or designee. Detailed closure requirements and completion dates or milestones for each b item were simultaneously provided. 48.13.1 Action to be completed prior to Restart. 48.13.1.1 Review the current Materials Management Program. CLOSURE REQUIREMENTS:
- 1. Establish a Materials Management organization.
O ( 2. Implement applicable sections of the LRS Report.
- 3. Identify elements of the current Materials Management program requiring modifications such as:
- Materials Requisitioning Process
- Materials Purchasing Process
- Procureirent Document Review
- Material Receipt Process
- Material Storage Practices
- Storage Maintenance Program
- Inventory Control Process
- Shelf-Life Program
- Materials Reservation System
- Materials Issue Process
- Resource Recovery Program
- 4. Prepare a flow chart of the current materials management process.
- 5. Implement the Materials Management Program in accordance with existing procedures during the period when new procedures are being prepared.
a 48.13-1
- 6. Prepare interim procedures and interim revisions to existing procedures as required to streamline the materials management process.
- 7. Conduct training to ensure that the Materials Management Program complies with the applicable requirements of the following documents:
- USAR
- Operations Quality Assurance Manual
- 10CFR50, Appendix 8
- 10CFR21
- ANSI N18.1 and its replacement ANSI 3.2
- ANSI 45.2.2 - 1972
- ANSI 45.2.13 - 1976 i - ANSI /ASME NQA 1979
- NRC Regulatory Guide 1.38 j - ASME/AWS I 8. Establish a Material Coordination Group.
- 9. Establish a Procurement Engineering Group.
f 10. Establish a Plant Procurement Group that is performing the functions of procurement, expediting, and traffic functions for procurement of items for Rancho Seco.
- 11. Augment the Material Control / Warehousing Group.
I
- 12. Provide on-the-job training on the interim Materials Management Program.
Completion date: Prior to release to Dispatcher. QTS reference: 25.0016 0 48.13-2
4C Plant Modifications and Maintenance improvements This Section consists of separate subsections for each Q system or area which is being mcdified, revised, or repaired before restart. Each subsection contains the related problem descriptions, a resolution statement, closure requirements, and forecast completion date or milestone. The resolution statements, closure requirements and completion date are provided by the assigned System Engineer. In some cases, the action actually being performed is seemingly different from the problem statement; however, the resolution statement is appropriate to resolve the original recommendation. The actions described in these subsections, therefore, comprise the plant modifications and maintenance work identified at this time to be done for restart. Actions which are not required for restart will be addressed in future amendments or by separate correspondence. 4C.1 Integrated Control System (ICS), Non-Nuclear Instrumentation (NNI) and Interfacing systems O- 4C.1.1 Problem 1 The Main Feedwater Pumps (MFPs) may contribute to an overcooling on a loss of ICS power. Tracking Number: 26.0037
Description:
On loss of ICS power, the MFPs will drop below minimum speed and can no longer be controlled from the Control Room. To meet the 10 minute goal for no operator action l after a reactor trip, the MFPs should be tripped so that the Once Through Steam Generator (OTSG) level will be controlled only by the backup Auxiliary Feedwater (AFW) controller. l Resolution: Modify the ICS to trip Main Feed Pumps on a loss of ICS power. Closure Requirement: ECN R-0823 Completion Date: Prior to loss of ICS power test at hot shutdown. O V l 4C.1-1 l _ _ . _. . _ . _ _ _ _ _ . _ . _ . ~ . _ _ _ _ . _ . _ _ _ ._ _ _ _ _ _ _ _ , _
4C.1.2 Problem 2 The ICS reacts erratically upon loss of NNI. Tracking Number: 26.0036
== Description:== The NNI provides 23 inputs to the ICS. These inputs to the ICS are the ICS " sensors" required for proper coordination of the turbine, feedwater and the reactor. Upon loss of NNI, a reactor trip will occur. Due to the complexity and variety of failure modes with the current design for loss of NNI, X, Y, or Z power, it would be unrealistic to expect an operator to sort out reliable controls and indication. Therefore, on loss of NNI, tripping the ICS and providing ICS and NNI independent Indications and controls will enable the operator to take the plant to a safe condition. Resolution: Modify the NNI to trip ICS power on a loss of NNI power. Closure Requirement: ECN R-0826 Completion Date: Prior to loss of ICS power test at hot shutdown. 4C.1.3 Problem 3 Procedures for restoration of ICS Power are not adequate. Tracking Number: 26.0038
== Description:== No procedures for restoration of or recovery from ICS power existed at the time of the transient. Resolution: Write procedures which address restoration of ICS power. Closure Requirement: Approval sheet of Casualty Procedure C.40 Revision 0. Completion Date: Complete. 4C.1.4 Problem 4 Electrical terminations in the ICS are inadequate. Tracking Number: 26.0039
== Description:== The primary cause of the 12/26/85 loss of ICS power was a changing high resistance connection in a lug crimp on the Positive 24 Volt Distribution Bus in ICS Cabinet #1 which connects this bus to the Positive 24 Volt Bus on the ICS Power Auctioneer Panel. 4C.1-2
Resolution: Inspect and repair electrical terminations in the ICS.
} Closure Requirement: Completed inspection of all terminations.
G Completion Date: Complete. 4C.1.5 Problem 5 The Power Supply Monitor (PSM) may have design flaw. Tracking Number: 26.0040
Description:
The Power Supply Monitor (PSM) Module measures the bus voltage on the same conductor which supplies current to the module. This makes the circuit extremely sensitive to any bus or connection resistance. Resolution: Study the design of the PSM (to be performed by outside laboratory). Closure Requirement: Copy of report from outside laboratory will be provided and work requests generated in response to negative findings therein. Completion Date: Complete. 4C.1.6 Problem 6 Annunciator procedures for ICS windows are inadequate. Tracking Number: 15.0004.8
Description:
During the review of the 12/26/85 trip sequence, a concern was raised that Annunciator Procedures for Window 34 "ICS Fuse Fall" and Window 64 "ICS or Fan Power Failure" on panel H2PSB were misleading and did not give enough guidance to the Control Room Operators. Resolution: Revise Annunciator Procedures to reflect changes in annunciation of ICS power. Closure Requirement: Procedures change approved by PRC. Completion Date: Complete. 4C.1.7 Problem 7 Auxiliary Steam Reducing Station control is not independent of ICS. 4C.1-3
Tracking Number: 15.0098
== Description:== Auxiliary steam pressure control station PIC-36014A failed on loss of power to ICS during the 12/26/85 trip. The result of this failure caused the auxiliary steam header to overpressure, lifting PSVs and adding to the depressurization of the main steam lines. Resolution: Modify auxiliary steam pressure station so that control is not lost on loss of ICS power. Closure Requirement: ECN R-0878 Completion Date: Prior to heat up. 40.1.8 Problem 8 The ICS Preventive Maintenance (PM) Program is not adequate. Tracking Number: 26.0047
== Description:== The following, as a minimum, should be added to the ICS PM Program:
- 1) Cabinet / module clean and inspect including fans and filters.
- 2) Procedures for module calibration.
- 3) Procedures for function testing.
- 4) Tuning ICS control loops.
- 5) Verification of Sl/S2 time delay and DC Power Supply capacity.
Resolution: 1. Write PM tasks for ICS cabinet / module cleaning and inspections.
- 2. Write PM tasks for ICS module calibration.
- 3. Write PM tasks to perform ICS function testing.
- 4. Write PM tasks to perform ICS tuning.
- 5. Write PM tasks to test or replace S1/S2 time delay switches.
- 6. Write PM tasks to test ICS power supply capacity.
Closure Requirement: PM Tasks to be developed and performed. Completion Date: Complete. 4C .1. 9 Pro'lem u 9 The ICS schematics and analog and digital drawings need corrections. Tracking Number: 26.0048 4C.1-4
)
Description:
ICS drawings N21.01 - 115 through 119 are extracted from B&W Instruction Book No. 620-0011. These drawings provide the analog logic for the ICS. Resolution: The ICS schematics and analog and digital drawings will be updated before plant restart. Also, the following B&W digital logic drawings will be incorporated into the site docussent control system:
- a. D 554914 ULD, digital logic 1
- b. D 554915 Integrated Master, Logic
- c. D 554915 Feedwater control digital logic Part I
- d. D 556492 Feedwater control, digital logic Part ll
- e. D 554917 Reactor Control, digital logic Closure Requirement: Orawing corrections completed.
Completion Date: Restart. 4C.1.10 Problem 10 Neutron error indicator from ICS cabinet H41002 to H1RI is not fused. Tracking Number: 15.0108
Description:
To improve the ICS power' reliability, all ICS AC loads external to the ICS cabinets should be fused. Resolution: Install a fuse (3/4 amp) on the neutron error signal line. Closure Requirement: ECN R-0442. Memo from System Engineer that all loads are now correctly fused. Completion Date: Complete. 4C.1.11 Problem 11 ICS AC power feeder breakers are two different amp ratings. Tracking Number: 15.0245
Description:
The AC feeder breaker (lJ04) to the ICS is 30 amps. The other AC feeder breaker (IC07) is 40 amps. 4C.1-5
Resolution: Replace the 30 amp breaker 1J04 with a 40 amp breaker. Closure Requirement: ECN R-0469 Completion Date: Complete. 4C.1.12 Problem 12 Power Supply Monitor (PSM) Sensing lines are " Daisy Chained" thereby degrading their effectiveness. Tracking Number: 15.0247-1
== Description:== The PSM measures the bus voltage on the same conductor which supplies current to the module. This makes the circuit extremely sensitive to any bus or connection resistance. Resolution: Modify the PSM circuit to sense directly off the power supply buses. Closure Requirement: ECN R-0359 Completion Date: Complete. 40.1.13 Problem 13 The "As Found" time delays for S1/S2 breakers were unacceptably different from design. Tracking Number: 15.0290
== Description:== During the ICS Equipment Investigation of the 12/26/85 Plant Trip, the ICS Shunt Trip Switches S1 and S2 were found to have a time delay of 0.144 seconds and 0.129 seconds. The Bailey technical manual for the monitor states that this delay is 0.5 seconds. The switch manufacturer stated that the switches should operate with a delay of between 0.2 and 0.8 seconds. The shorter time delay makes the shunt trip sensitive to short duration transients which would not otherwise affect the ICS power supplies. Although this in itself could not cause this trip, it increased the probability of it occurring. Resolution: Test and install new S1 and S2 shunt trip switches in ICS and NNI. Closure Requirement: Work Request completion. Completion Date: Prior to loss of ICS test at hot shutdown. O 4C.1-6
4C.1.14 h Q Problem 14 During a " loss of ICS", plant trip operator action is required immediately and constantly to prevent ADV control of steam header pressure. Tracking Number: 21.0103
Description:
Operator experience at the simulator in Lynchburg has , Indicated that maintaining steam header pressure using the present setup (a switch which provides full closed and 507. open positions) requires the full attention of two licensed operators during a " loss of ICS" plant trip. A distraction could easily lead to overcooling or EFIC opening the ADVs, creating the potential for an s uncontrolled release. Resolution: Modify TBV controls to provide remote full throttle control independent of ICS on loss of ICS DC power. Closure Requirement: ECN R-0861 Completion Date: Prior to leaving cold shutdown. 4C.1.15 Problem 15 Main Feed Pump minimum speed is too low. Tracking Number: 26.0940
Description:
MFP minimum speed is set at 2600 RPM instead of 2800 RPM. Resolution: Reset the Main Feed Pumps minimum speed to its original setting of 2800 RPM. l Closure Requirement: Work Requests 119390 & 124977 Completion Date: Prior to leaving cold hutdown. 4C.1.16 Problem 16 The ICS/NNI Power Supply reliability needs improvement. Tracking Number: 23.0013-1
Description:
Currently the ICS is supplied by vital power from the C inverter (IC07) and also from a non-vital source - the J inverter (lJ04). The NNI is supplied from the D inverter (1007) and from the J inverter (1J08). The current Inverter design and arrangement has caused several Reactor Trips. k 4C.1-7
Resolution: Change power sources of the ICS and NNI from the C and J inverters to S1GA and S1G8 inverters. Closure Requirement: ECN R-0927 Mod 81. Completion Date: Prior to leaving cold shutdown. 4C.1.17 Problem 17 ICS Function Generator Modules load NNI signal source when ICS power is lost. Tracking Number: 26.0027
== Description:== Function Generator Modules (6624665-1) do not have adequate input impedance when the 24 volt supply is removed (i.e., loss of ICS DC power). Resolution: Modify the Function Generator Modules to prevent loading its inputs. Closure Requirement: ECN R-1217, complete. Completion Date: Prior to cold shutdown functional test of ICS. 4C.1.18 Problem 18 Power Supply Monitor (PSM) does not directly sense the auctioneer bus. Tracking Number: 15.0126
== Description:== The PSM measures the bus voltage on the same conductor which supplies current to the module. This makes the circuit extremely sensitive to any bus or connection resistance. Re.2olut ion: Modify the PSM circuit to sense directly off the power supply buses. Closure Requirement: ECN R-03598, complete. Completion Date: Complete. 4C.1.19 Problem 19 No procedures exist for loss of signal conversion cabinets. Tracking Number: 26.0050 4C.1-8
1 i
Description:
The signal conversion cabinets provide the power for the ! g signal developed by several transducers. The signal is
'\ then supplied as input to the NNI System.
Resolution: 1. Write procedures to deal with a loss of the Signal Conversion Cabinets A and B.
- 2. Train operators on actions required upon loss Signal Conversion Cabinets A and B.
Closure Requirement: Approval sheet of procedure (to be supplied). Lesson plan for loss of Signal Conversion Cabinets A and B. Letter from training stating that the training has been given to all Control Room operators. Completion Date: Prior to heat up. 4C.1.20 Problem 20 Blas Module Circuit Boards are discolored due to heat. The bias voltage drifts. Tracking Number: 26.0029
Description:
The ICS cabinets at Davis-Besse are provided with fewer cooling fans than the Rancho Seco ICS Cabinets. For
/ this reason, the Blas circuit drift problem was (7s
) accelerated at Davis-Besse. Several modules with Blas circuitry have been changed out at Davis-Besse.
Resolution: Upgrade the bias modules to prevent overheating. Closure Requirement: Work Request 109623 complete. Completion Date: Prior to cold shutdown functional test of ICS. 4C.1.21 Problem 21 AC fuses for ICS loads may be oversized. Tracking Number: 26.0053
Description:
The ICS AC fuses may be oversized and could therefore contribute to an ICS loss of power. Resolution: Change the AC fuse sizes for ICS loads to .75 amps. Closure Requirement: ECN R-0927. Completion Date: Prior to heat up. O 4C.1-9
4C.1. 22 Problem 22 Procedures are inadequate for loss of NNI power (Hot S/D to Cold S/D). Tracking Number: 18.0025-1
== Description:== As a result of the Deterministic Failure Analysis on ICS/NNI, modifications to the ICS/NNI will require a rewrite of existing loss of NNI procedures. Resolution: Revise loss of NNI procedures. Closure Requirement: Approval sheet of procedures. Completion Date: Prior to heat up. 4C.1. 23 Problem 23 Annunciation of ICS power status is ambiguous. Tracking Number: 15.0004A
== Description:== During the review of the 12/26/85 trip sequence, a concern was raised regarding the possibility that the combination of ICS power failure and cabinet fan failure in one annunciator window could mislead the operator, or extend the time taken to respond. Hence, an evaluation of ICS trouble annunciation was requested. Resolution: Modify ICS/NNI power annunciation scheme to clarify a true power failure in the ICS. Closure Requirement: ECN R-0517, complete. Completion Date: Complete. 4C.1. 24 Problem 24 Annunciation windows for NNI power status are ambiguous. Tracking Number: 23.0012
== Description:== During the review of the 12/26/85 trip sequence, a concern was raised regarding the possibility that the combination of ICS power failure and cabinet fan failure in one annunciator window could mislead the operator, or extend the response time. Hence, an evaluation of ICS trouble annunciation was requested. Due to the slmtlarity of ICS and NNI, the request for evaluation was broadened to include the NNI annunciation. 4C.1-10
Resolution: Modify ICS/NNI power annunciation scheme to clarify a true power failure in the NNI. b Closure Requirement: ECN R-0580, complete. Completion Date: Complete. 4C.1.25 Problem 25 Status of the ADVs, TBVs, main feedwater and start up feedwater valves position is not indicated in the Control Room. Tracking Number: 26.0033
Description:
Status of these valves is currently not provided in the Control Room. Since these valves have the potential to fall in a position which may cause an overcooling event, their position is important information that should be available to the operator in the Control Room. Resolution: Install indicating lights in the control room which will provide status of these valves. Closure Requirement: ECN R-0828 Completion Date: Prior to heat up. 4C.1.26 Problem 26 Loss of ICS Procedures and Training are not adequate. Tracking Number: 26.0035
Description:
Additional concerns, not addressed in current procedures or training, were revealed during the DFC and other analyses performed on the ICS. Also, as a result of the modifications to be installed prior to start up, modification training and procedure updates need to be done. Resolution: 1. Revise procedures to reflect loss of ICS.
- 2. Revise operator training to reflect loss of ICS.
Closure Requirement: New procedures accepted and approved. Lesson plans for training on loss of ICS. A letter from training stating that loss of ICS training has been given to all control room operators. Completion Date: Prior to heat up. 4C.1-Il
I 4C.1.27 Problem 27 Annunciator procedures for loss of NNI are inadequate. Tracking Number: 15.C J1 l l
== Description:== Dur j the review of the 12/25/85 trip sequence, a i conturn was raised that Annunciator Procedures asscciated with the ICS power failures were inadequate and ineffective. The concern was broadened to include , anntaciators associated with NNI power failure. l Resolution: Revise annunciator procedures for loss on NNI. Closure Requirement: Approval sheet for new procedure (to be supplied). Completion Date: Conplete. 4C.1.28 Problem 28 Additional training is needed for NNI power failure. Tracking Number: 18.0026
== Description:== Casualty Procedures for loss of NNI (C.14, 15, 16, 17) will be revised before plant start up. In addition, plant modifications associated with loss of NNI will be made. Operators will need to be trained on these procedure changes and modifications. Resolution: Train operators on modifications and procedure changes relating to loss on NNI power. Closure Requirement: Lesson plan for loss on NNI power. Memo from training starting that the training has been given to all Control Room operators. Completion Date: Prior to heat up. 40.1.29 Problem 29 All the Safety Parameters Display (SPDS) inputs are not independent of NNI Power Supply. Tracking Number: 22.0005
== Description:== The SPDS is not independent of NNI (to get to Hot Shutdown). Some of the inputs to SPOS are provided by NNI, making it vulnerable to a loss of NNI power. A list of the signals was given to the Nuclear Engineering ILC Group. 4C.1-12
Resolution: Modify the input power supplies to SPOS of those ~ channeIs needed for hot shutdown to separata them from NNI. Closure Requirement: ECN R-0953. Completion Date: Prior to heat up. 4C.1.30 Problem 30 Electrical terminations in NNI may be unreliable. Tracking Number: 26.0076
Description:
The primary cause of the 12/26/85 loss of ICS power was a changing high resistance connection in a lug crimp on the Positive 24 Volt Distribution Bus in ICS Cabinet #1 which connects this bus to the Positive 24 Volt Bus on the ICS Power Auctioneer Panel. The NNI electrical terminations should also be inspected. Resolution: Inspect and repair electrical terminations in the NNI. Closure Requirement: Work Requests (to be supplied), completed. ECNs (to be supplied), completed. NCRs (to be supplied), completed. Completion Date: 2/1/86 4C.1.31 Problem 31 NNI X negative 24V DC power supply does not meet manufacturer's specifications for AC ripple and noise. Tracking Number: 26.0077
Description:
NNI X Negative 24V DC power supply does not meet manufacturer's specification for AC ripple and noise. This could possibly cause the power supply not to perform as required. Resolution: Rework the power supply. Closure Requirement: Work Request 113849. Completion Date: Complete. 4C.1.32 Problem 32 Rewrite procedures for restoration of NNI power. 4C.1-13
Tracking Number: 23.0033
== Description:== Procedures for restoration of NNI power need to be updated. Resolution: Revise procedures for restoration of NNI. Closure Requirement: Approval sheet for casualty procedure C.15. Completion Date: Prior to heat up. 40.1.33 Problem 33 Review of open NNI Work Requests (WRs) (prior to 7/12/86). Tracking Number: 26.0105
== Description:== The Plant Maintenance Daily Work Listing (dated 07/12/86) was reviewed for open work items that must be completed prior to start up. Resolution: Complete work requests listed below prior to restart. Closure Requirement: Work Request 110365, 113769, 113770, 113771, 113772, 113774, 113777. Completion Date: Prior to restart. 40.1.34 Problem 34 Review of Open ECNs. Tracking Number: 26.0395
== Description:== The list of open ECNs for the NNI System was reviewed. The list below contains open NNI related ECNs that are not part of this report or are not covered by another System Status Report. Resolution: Close the ECNs listed below prior to restart. Closure Requirement: ECN A-5620A through F, A-5760, A-5249, R-0499, R-0594, R-0685, R-0919. Completion Date: Prior to restart. O 4C.1-14 i
4C.1.35 Problem 35 Review of Open NCRs. s.s Tracking Number: 26.0396
Description:
The list of open NCRs for the NNI System was reviewed. Resolution: Close the NCRs listed below prior to restart. Closure Requirement: NCR S-5302, S-5381, S-5386, S-5429, S-5466, S-5468, S-5486, S-5487, S-5543, S-5546, S-5710. Completion Date: Prior to restart. 4C.1.36 Problem 36 All fuses on the Z-24V DC bus are not alarmed in the Control Room. Tracking Number: 26.0397
Description:
Due to four relays not installed on the Z bus fuse panel, the fuses on this panel will not alarm in the Control Room. Resolution: Modify the NNI so that fuses on the Z DC bus will alarm Q in the control room. Closure Requirement: ECN R-1255 Completion Date: Prior to heat up. 4C.1.37 Problem 37 Loss of one signal conversion cabinet could cause dryout of both once Through Staw Cenerators (OTSGs) if plant trips while start-up levr's were converted by the same cabinet. Tracking Number: 22.0593
Description:
Loss of one signal conversion cabinet could cause dryout of both OTSGs if plant trips while start-up levels were converted by the same cabinet. Resolutlon: Instali EFlC. Closure Requirement: Completion of ECN 5415 (All). Completion Date: Prior to restart. 4C.1-15
4C.1.38 Problem 38 Actions requested or required by two 1980 NRC letters could not be verified as completed. Tracking Number: 20.0382
== Description:== The 03/06/80 NRC letter to all B&W licensees and 04/14/80 letter transmitting a Confirmatory Order from the NRC deal with loss of power to ICS and/or NNI. Actions requested or required of the two letters could not be verified to have been completed. Resolution: Verify that the Shutdown Panel indicators at H4BS are independent of NNI (loss of ICS/NNI test). Closure Requirement: Approval sheet for procedure. Completion Date: Loss of ICS/NNI test at hot shutdown. 4C.1.39 Problem 39 The HPl flow meters indications are affected by NNI AC or DC power supply failures. Tracking Number: 26.0459
== Description:== The HPI flow meters receive input from the NNI system and are directly impacted by various NNI AC or DC power supply failures. The operator can be confused when attempting to evaluate the performance of this safeguard system following SFAS initiation as a result of, or subsequent to loss of NNI AC or DC power. Resolution: Modify the HPl flow indications so that they are independent of NNI. Closure Requirement: ECN R-1272, completed. Completion Date: Prior to heat up. 4C.1.40 Problem 40 Incomplete IE Notice 84-80 Actions. Tracking Number: 20.0495 0 4C.1-16
Description:
IE Notice 84-80 dealt with the partial loss of NNI power at Rancho Seco on 03/19/84, and the loss of NNI Y power at Crystal River 3 on 04/26/84. During the Precursor Review process, it could not be verified that actions described in the notice had been completed. ; l Resolution: After verification this item can be closed.
- a. ECN A-5112 was closed on 06/08/85.
, b. Installation of the permanent multi-point recorder installed per ECN R-0876.
- c. PM Tasks 4146 (NNI) and 4145 (ICS) were written to check the 1 24V DC power supply alarm and trip I
- settings. '
- d. NNI drawings and plaques on the NNI panels were updated and replaced per WR 90466.
- e. Plant training manuals were updated per CCL No.
, 84-0455. (Closed per Jack Nau meno to Steve Crunk ! dated 09/26/84, Jp0184-106.)
- f. ECN A-5318 was written to correct power supply
- wiring in the ICS and NNI panels. This ECN was l volded by NCR S-3779. Work Request 85883 provided the corrective action for NCR S-3779.
Closure Requirement: Complete. j Completion Date: Complete. l l I l l t l
\
4C.1-17
l 4C.2 Auxiiiary Steam System (ASC) b 4C.2.1 Problem 1 Wiring hazards may exist in Auxiliary Boiler Panel H2A8. Tracking Number: 26.0610
Description:
Hiring in Auxiliary Boller Panel H2AB is not consistent with accepted wiring standards. Resolution: Modify panel H2A8 wiring to conform to standards. Correct electrical drawings to reflect actual configuration. Closure Requirement: ECN A-5765 NCR 2-4781 Completion Date: Prior to condenser vacuum. 4C.2.2 Problem 2 Air / Fuel ratio gauge on Auxiliary Boiler E-360 is not functioning properly. Tracking Number: 26.0252
Description:
Auxiliary Boiler E-360 Air / Fuel ratio controller HIK-36101B is not operating properly. Resolution: Repair and recalibrate HIK-361018. Closure Requirement: HR 118653 Completion Date: Prior to condenser vacuum. 4 4C.2.3 I Problem 3 Boiler burner modifications are required. Tracking Number: 26.0255
Description:
Present burner on Auxiliary Boiler E-360 is inefficient and produces a poor flame pattern. Current boiler configuration provides too much flame on low flame setting. , Resolutlon: InstalI new burner in boller E-360. ) 4C.2-1
, - - - + - , - -
m,., _.m -, ._.---,yp, g- ,,g.,_ , , , ,, p cy_,gy. _ . , y..,. ...e,w,...,_,-----ww-,,.g,,. .=%.,,9p.y 3,, - ,+,,m,,7-.-_.,m,, .g ,-_%.,.-,,-pp. ,
Closure Requirement: ECN R-1218 HR 118091 Completion Date: Prior to condenser vacuum. 4C.2.4 Problem 4 Review of Priority 1 open ECNs. Tracking Number: 26.0243
Description:
Boller startup circuitry requires at least a 107. fuel oil flow setting for light-off. Resolution: Revise Operations Procedure A.39 to ensure adequate safety precautions concerning fuel oil flow setting are followed during boiler startup. Closure Requirement: Procedure A.39 revision. Completion Date: Prior to condenser vacuum. 4C.2.5 Probles 5 Procedure A.39 needs to be rewritten. Tracking Number: 22.0101
Description:
Enclosure 8.1 of A.39 does not contain the valve PV-35048 and its associated manual valves, and their required position. Resolution: Revise Operations Procedure A.39, Enclosure 8.1 to add , valves PV-35048, ASC-610, ASC-011, ASC-652, ASC-777, ASC-479, and ASC-480. Closure Requirement: Procedure A.39 revisions. Completion Date: Prior to condenser vacuum. 4C.2.6 Problem 6 Auxiliary boiler should have a control system tuneup. Tracking Number: 26.0246
Description:
A tuneup of the auxiliary boiler control system should be performed. Resolution: Complete an auxiliary boiler control system tuneup on both auxiliary boilers. l 4C.2-2
Closure Requirement: WR (later)
/ Completion Date: Prior to condenser vacuum.
C/ 4C.2.7 . Problem 7 Review of open ASC Work Requests (as of 08/21/86). Tracking Number: 26.0247
Description:
- 1. HR #92436 - Insulation is required on line 61250-1-GB in the Tank Farm area.
- 2. WR #107127 - Shaft guards are required on boiler forced draft fans A-361 and A-366.
- 3. HR #117124 - Boiler E-365 vent valves FV-36543 indicates intermediate position with valve in closed position.
- 4. HR #117492 - Auxiliary Steam supply header relief valve PSV-36012A requires setpoint determination and adjustment.
- 5. HR #117498 - Alarm horn outside of boiler shack does not sound upon receipt of alarm at panel H2AB.
Resolution: Complete Work Request 92436,107127,117124,117492, and 117498. Closure Requirement: hrs 92436, 107127, 117124, 117492, 117498
, Completion Date: Prior to condenser vacuum.
l-O 4C.2-3
4C.3 Emergency Feedwater Initiation and Control (EFIC)
.Ch b 4C.3.1 Problem 1 EFIC/M00 1 is not installed.
Tracking Number: 26.0706
Description:
NUREG 0578 and NUREG 0737 formed the basis of an upgraded AFW initiation and control system. The
" Emergency Feedwater Initiation and Control" (EFIC)
System provides: a) Redundant automatically initiated AFW for all AFW design basis events, b) Redundant safety giade control of AFW to assure sufficient, but not excessive AFW flow. c) Isolation of MFW and AFW to prevent continued feed to an OTSG in the event of a SLB inside containment. d) Fa11 safe control of ADVs with diverse power sources for ADV control circuits. Resolution: Install EFIC/M001 with the exception of sub-ECNs A-5415 R, T, and AF. Closure Requirement: Install EFIC/ MOD 1 with the exception of sub-ECNs A-5415 R, T, and AF. Completion Date: EFIC Actuation Test. I L l 'O i 4C.3-1 i
- _ , . . . _ . . . , . . . - . , , - . _ _ - _ _ _ _ _ . - _ _ . . . . _ . , _ _ , . _ _ _ _ _ _ _ _ _ . , _ _ _ _ . . _ _ - , _ .4._..
4C.4 120VAC Electrical System (120VAC) V 4C.4.1 Problem 1 120VAC Vital Bus unreliability. Tracking Number: 26.0267
Description:
The original 120VAC vital inverters have proven to be unreliable. Failures of these inverters have resulted in several reactor trips. Resolution: Regulating transformers will be installed to provide an alternate power source to the 120VAC vital buses. The inverters will provide the normal power to the 120VAC vital buses. Closure Requirement: ECN-R-0955 Completion Date: Loss of Site Power Test 4C.4.2 Problem 2 Review of priority one open Work Requests. Tracking Number- 26.0264
Description:
The open work requests required to be completed to ensure system operability were identified. Resolution: The work requests shall be completed prior to restart. Closure Requirement: Close W/R law quality procedures. Completion Date: Loss of Site Power Test 4C.4.3 Problem 3 Update the NRC IE Bulletin 79-27 study. Tracking Number: 26.0353 1
Description:
The present study for IE Bulletin 79-27 was completed in 1980. This study dealt with the loss of instrument and control buses during operation. Since that study was completed, there have been numerous changes to the 120VAC vital buses. Resolution: lbdate the IE Bulletin 79-27 study to include all cianges to the 120VAC vital buses. 4C.4-1
Closure Requirement: Nuclear Engineering activity. Completion Date: Instrument Power Test 40.4.4 Problem 4 NCR S-4723 disposition and ECN A-5710 do not agree. Tracking Number: 26.0938
== Description:== The disposition of NCR-S-4723 does not agree with the engineering change recommended in ECN-A-5710. Resolution: N.E. to evaluate disposition of NCR-S-4723 and revise ECN to provide agreement between the documents. Closure Requirement: Nuclear Engineering Manager memo stating satisfactory resolution. Completion Date: Prior to performance of work activity for HR's #117228, 117229, 117230, and 117231. 1 O O 4C.4-2
4C.5 Containment Building Spray System (CBS) \ 4C.5.1 Problem 1 Reactor Building spray inlet valves need to be provided with modulating control. / Tracking Number: 19.0009
Description:
Remove " seal-in" contacts both open and closed directions, for the reactor building spray inlet valves SFV-29107 and SFV-29108. This will to provide modulating control in manual operation for these valves. These valves must still stroke full open on - SFAS. This change was previously proposed on ECN R-0704 and volded per memo from Tar Singh to D. Cameron on 4 06/10/86. Office Memorandum NOM 86-338 dated June 5, 1986, from Steve Redeker to John Jewett, RRRB Chairman, and Tom Tucker, Operations Superintendent, stated that ' Engineering has concluded that the Reactor Building Spray Valves do not need to be modulating because the . pumps would not "run out" due to system configuration. The recommendation to make valves prior to plant start up is withdrawn, and should be removed from consideration by the RRRB. , , Resolution: Operating procedures will be revised on criteria of high flow for Cas pumps. Closure Requirement: Revision of A.7, the associated annunciator procedure, process standard (AP.105) and associated response procedure. The alarm set point revised and instrument recalibrated. (FSGK-29103 and 29104) to reflect new set point. Completion Date: Prior to restart (criticality). 4C.5.2 Problem 2 Review of Open CBS Work Requests (prior to August 12, 1986) Tracking Number: 26.0205
Description:
Plant Maintenance Daily Work Listing (dated 8/12/86) review for open work items that must be completed prior to start up.
- 1. WR# 116980, Safety Features Valve SFV-29015, spare O conductors were found wrapped with duct tape instead of approved Ray Chem WCSF Tubing. As a result NCR S-5600 was written.
4C.5-1
f
/*
- 2. Ammeter for P-291A located on H2SFA (Control Room),
suspected broken as per NCR S-5645. HR# 117258 provides inspection and repair, if required.
- 3. Reactor Building Spray Pump P-2918 has vibration 4 readings in the " Alert Range." WR# 097873 and
, 2
#,. 099391 double frequency on pump readings, check alignment and forward results to Engineering for evaluation. An NCR is not warranted at this time as an additional evaluation from Engineering is not yet required.
- 4. The following work requests consisted of PMs on snubbers.
115454 - Snubber #45W-29101-1A 115457 - Snubber #1SH-29125-7A 115461 - Snubber #1SW-29122-28A 115466 - Snubber #4SH-29101-1A
, 115469 - Snubber #1SH-29125-1A 115473 - Snubber #1SH-29122-1A
- 5. Reactor Building "B" Spray Pump Relief Valve l PSV-29118 had exceeded the calibration period as I
., specified by Surveillance Procedure SP 214.02 and ASME Section XI, IWV-3510. Verify set point per SP
, 214.02 and reset as necessary.
Resolution: Complete listed open work requests. Engineering will )
/ evaluate vibration on P-2918 and implement any action. I Closure Requirement: Completed work requests: 116980, 117258, 097873, l 115454, 115457, 115461, 115466, 115469, 115473.
NCR 55645, NCR 55600, and surveillance procedure SP 214.02. Evaluation on CBS pump (P-2918) and ; d corrective action.
'l Completion Date: Prior to restart (criticality).
l [ O 4C.5-2
4C.6 Borated Water System (BWS) 4C.6.1 Problem 1 BWS (Borated Water System) NCR to be closed prior to start up. Tracking Number: 26.0642
Description:
The Injection Header Warming Pump shaft continued to bind up after maintenance. This pump is needed to prevent boric acid precipitation in the lines between the Decay Heat Removal Pumps and the BNST. b Resolution: Mechanical Maintonacce will complete WR 113141 and closing out NCR S-5912 on pump. Closure Requirement: HR# 113141 and NCR S-5912 closure. Completion Date: Prior to restart (criticality). _ [ 4C.6.2 Problem 2 Work Request to be closed prior to start up (BWS). Tracking Number: 26.0643
Description:
The following Work Requests on the BHS should be completed and closed prior to start up. Resolution: Maintenance shall complete and close out work request. Closure Requirement: Complete WR# 119866, 119867, 123131, 123132, 117157, and 111510. Completion Date: Prior to restart (criticality). O 4C.6-1
l 4C.7 Core Flood System (CFS) p 4C.7.1 Problem 1 Procedure needs to be revised. Tracking Number: 26.0288
Description:
During a check by Operations, the temperature of the "B" Core Flood Tank was 59'F and pressure was 560 psig. Operating Procedure A.4, limit and Precaution 3.6 states, "If tank temperature is less than 70*F, maximum allowable pressure is 140 psig." Revise the following procedures by the indicated department:
- 1. OP A.4 - Delete reference to temperature. Add:
The CFTs shall be depressurized whenever the plant is brought to cold shutdown. Nuclear Plant.
- 2. NEP 5409, " System Design Basis-Core Flood System (CFS)." Add: Hydrotests required by future modifications shall be performed at a temperature of at least 96*F. Nuclear Engineering.
O 3. AP.27, "Blannual Procedure Reviews" - Revise review V cover sheet to include a review of the operating procedures against the pertinent system design basis. Nuclear Plant. Resolution: Revise procedures OP A.4, NEP 5409, AP.27 and close out NCR S-5389. Closure Requirement: Complete Procedures modification and close out NCR. Completion Date: Prior to restart (criticality). 4C.7.2 Problem 2 Priority One Open CFS Work Requests (Prior to 9/2/86) Tracking Number: 26.0289
Description:
Plant Maintenance Daily Work listing was reviewed for open Work Requests that must be completed prior to start up.
- 1. WR# 107034 - Test PSV-26510 per SP 214.02 (NCR S-5257).
- 2. WR# 117624 - Fix pipe support lu-26528-14 (NCR S-5784).
4C.7-1
- 3. HR# 117654 and 115405 - Replacement of snubber seals is to be performed in response to the snubber seal life program (NCR S-5656).
- 4. HR# 117682 and 115507 - Replacement of snubber seals is to be performed in response to the snubber seal life program (NCR S-5656).
- 5. HR# 110490 - Reca11brate LT-26505.
- 6. HR# 117538 - Computer point (L9004) for "B" CFT Level alarms on and off on the IDADS typewriter.
- 7. HR# 112577 - HV-26517 needs additional packing, only 1/8" of adjustment left.
Resolution: Complete listed open Work Requests. Closure Requirement: Complete the following Work Requests: 107034, 117264, 117654, 117682, 110490, 117538, 112577, NCR S-5257, S-5784, S-5656 surveillance procedure SP 214.02. Completion Date: Prior to restart (criticality). 4C.7.3 Problem 3 "A" Core Flood Tank Vent Valve HV-26511 Leaks. Tracking Number: 26.0291
== Description:== In 1977, a Work Request was written to repair the valve seat for HV-26511. The seat could not be repaired in place. A blank flange was installed in the flow orifice downstream of the valve with the intent to replace the valve during the following refueling outage. The Work Request was closed out. In 1982, a Work Request (HR# 064682) was written to repair or replace the valve, remove the blank flange, and to verify there is not a blank flange in the "B" Core Flood Tank vent line. While walking down the vent line portion of the system, the following was discovered:
- 1. There is not a blank flange on the "B" Core Flood Tank line.
- 2. he h a blank flange on the "A" Core Flood Tank 4C.7-2
- 3. There are no tags on the "A" vent line, HV-26511, (q
FG-26519, or CFS-015 indicating that there is a blank flange installed or CFS-015 has a potential of having 600 psig behind it with HV-26511 closed. This could be a safety hazard. (See ODR# 86-61.) Resolution: 1. Mechanical maintenance to remove blank flange from FO-26519.
- 2. Nuclear Engineering to purchase acceptable substitute valve and evaluate cutting and capping the vent line downstream of the flow orifice on both tanks.
Closure Requirement: HR# 064682 closed out. Completion Date: Prior to pressurizing the core flood tank. 4C.7.4 Problem 4 FulI Stroke Testing of CFS-001 and CFS-002. Tracking Number: 26.0310
Description:
Check valve CFS-001 and CFS-002 are not currently under IST Program as required by Section XI and ASME. 1974 edition summer 1975 agenda. Review of NRC letter dated September 25, 1984, addressed to SMUD, Docket No. 50-312, revealed that the NRC has disapproved the relief request or partial stroke test of the check valves CFS-001 and CFS-002. The relief was only allowed up to September 25, 1985. The NRC has stipulated that after this expiration date of relief, the licensee (SMUD) must demonstrate the capability of each valve to actuate to at least the position required to fulfill its safety function. Resolutlon: Perform a study to determine the proper test method and/or any design modifications required to accomplish the test program. A test program shall be developed and implemented prior to start up to verify that the valves will perform their safety function. Write Surveillance procedures to test valves CFS-001 and CFS-002. Closure Requirement: Issue new SP 91. Completion Date: Prior to placing CFS back in service. O 4C.7-3
4C.8 Radiation Monitoring System (RMS) 3 4C.8.1 Problem 1 Recorder RJR-013 cannot be easily read. Tracking Number: 26.0284
Description:
Radiation Monitor Recorder RJR-013 does not differentiate between the various monitor points to allow assessment of radiological releases and conditions before, during, and after an event. Resolution: Closure of ECN R-0457 and WR# 115507. Cicsure Requirement: Closure of ECN R-0457 and HR# 115507 Completion Date: Prior to restart. 4C.8.2 Problem 2 Discrepancies between R-150028 and R-15045. Tracking Number: 15.0011 At the beginning of each year, new set points are l O
Description:
calculated for the auxiliary building vent monitors using a default nuclide mix and the latest annual average dispersion coefficients (X/Q). Since the set point calculations are based on the same default nuclide mix and utilize the same X/Q and flow rate, the set points should be the same. Resolution: Implement new set points. Closure Requirement: Revision to the off-site dose calculation manual and process standards. I . Completion Date: Complete. 4C.8.3 Problem 3 Correlate R-15049 and R-15050 with Reactor Building activity. l Tracking Number: 15.0019
Description:
Data is needed for a back-up on the PASS system to l correlate isotopic mix with core damage and Reactor
/ Building gaseous activity.
l O] 4C.8-1 l
Resolution: Revise Emergency Planning procedure, AP.511, to make use of the Engineering curves developed. Closure Requirement: Approval sheet of AP.511 revision. Completion Date: Prior to restart. 4C.8.4 Problem 4 R-15017A&B not in service. Tracking Number: 21.0111
Description:
Liquid Effluent Monitors R-15017A, B were installed about two years ago but have not been placed in service yet, due to plugging and in ability to maintaining flow. Resolution: Install submersible pumps. Closure Requirement: Closure of ECN A-4714 and WR 84219, 84222, and 115150. Performance of STP-648. Completion Date: Prior to start up. 4C.8.5 Problem 5 R-15001 and R-15002 - multiple problems. Tracking Number: 21.0110
Description:
Multiple problems with R-15001 and R-15002 Resolution: Repair and upgrade monitors or replace. Closure Requirement: Completion of ECNs A-4052, R-0783, R-0228, and WR 114845 and 114846, or if monitors are replaced, completion of ECN R-0913. Completion Date: Prior to restart. 4C.8.6 Problem 6 R-15020 failure rate is high. Tracking Number: 21.0113
Description:
Regenerant Hold-up Tank Monitor R-15020 fails often due to plugging with resin. Resolution: Provide filtration prior to radiation monitor and a monitor flow sample point. I 4C.8-2
Closure Requirement: Completion of ECN R-0891 and closure of associated NCR S-5156
'V Completion Date: Prior to restart.
4C.8.7 Problem 7 Review of Open ECN R-0295. Tracking Number: 26.0276
Description:
Open ECNs to be closed prior to restart. Resolutlori: Complete work. Closure Requirement: Closure of ECN R-0295. Completion Date: Prior to start up. 4C.8.8 Problem 8 WRs to be completed prior to restart. Tracking Number: 26.0278 y]
/
Description:
The Plant Maintenance Daily Work Listing was reviewed for open work items that must be completed prior to restart. Resolution: Complete work prior to restart. Closure Requirement: Closure of the following work requests: 112122 Reinstall detectors and calibrate R-15044 112123 Reinstall detectors and calibrate R-15045 112124 Reinstall detectors and calibrate R-15046 115983 Calibrate R-15001C when clearance is removed. 113635 R-15001D would not respond to inputs 11x10(4) CPM 113758 Resolve discrepancy between area monitor reading and recorder R-15041 115995 Perform I.605 (calibration) when clearance removed (R-15001A) 117483 R-15006 Flow ratemeter pegged high 101708 R-15546 Remove abnormal tag 3836 106317 R-15049 Failed source check 107962 R-15045 Continual spiking Completion Date: Prior to restart. - O 4C.8-3
4C.8.9 Problem 9 Open NCRs to be completed prior to restart. Tracking Number: 26.0280
== Description:== NCRs determined to be priority 1. Investigation: S-4845 Check source response during surveillance not readable due to high background - replace sources with high activity sources, HR 117212, also 113632, and 113633. S-5608 Control Room H4HRA Rad Monitor panel overheats. ECN R-0872 to provide cooling fan. Resolution: Complete disposition of NCRs prior to restart. Closure Requirement: Closure of NCR S-4845 and S-5608 upon completion of associated ECNs and hrs. Completion Date: Prior to restart. 40.8.10 Problem 10 Technical Specification statement requires resolution. Tracking Number: 27.0042
== Description:== Resolve statement in " Discussion and Evaluation" section of Amendment 73 of the Tech Specs that Rancho Seco does not have "Same as channels which can be calibrated" when we do. Investigation: An incorrect statement was made by the NRC in Amendment 73 to the Technical Specifications regarding Rancho Seco not having "same as channels which can be calibrated" when in reality it does. This was discussed with the NRC. A letter to the NPC was prepared but not sent. Resolution: Provide resolution prior to restart. Closure Requirement: Memo from Licensing. Completion Date: Prior to restart. 4C.8.11 Problem 11 Main Steam Line Monitor readout. Tracking Number: 22.0389 4C.8-4
Description:
R-15047 and R-15048, Main Steam Line Monitors. only es l reads out at the Control Room console HlDRMS which is powered from "G" Inverter (72-E27). Tech Specs Table (d 3.5.5-1 requires both channels operable. ResoIution: Compieta modification and close out ECN R-1041. Closure Requirement: Complete ECN. Completion Date: Prior to restart. 4C.8.12 L Problem 12 Inadequate RCS leak detection. Tracking Number: 26.0283
Description:
Tech Specs require Reactor Building Particulate Monitor for RCS leak detection. Resolution: 1. Create an additional leak detection method.
- 2. A Technical Specification change.
Closure Requirement: 1. New leak detection method.
- 2. Tech Specs change approved.
Completion Date: Prior to restart. 4C.8.13 Problem 13 Determine use and setpoints for R-15001. Tracking Number: 27.0063 -
Description:
PRC Minutes of 11/04/85 questioned the Technical Specification requirements on the use of R-15001. Licensing was assigned to resolve its use and setpoints. Resolution: Determine use and setpoint criteria for technical specification requirements. Closure Requirement: Implementation of new setpoint criteria. Completion Date: Prior to restart. 4C.8-5
4C.9 Safety Features (SFS) 4C.9.1 Problem 1 Electrical terminations in the SFAS cabinets need to be reworked. Tracking Number: 26.0139
Description:
Deficiencies have been found in the SFAS cabinets electrical terminations. Resolutlon: Repair / rework alI electrical termination deficiencies. Closure Requirement: Closure of the following: NCR ECN S-5371 R0585 S-5375 R0584 S-5378 R0595 S-5387 - S-5397 S-5399 - S-5410 R0588 S-5418 -----
' S-5435 S-5492 -----
S-5495 R0637 S-5510 S-5596 R0830 S-5612 R0587 Completion Date: Prior to functional testing of SFAS in cold S/D. 4C.9.2 i Problem 2 Procedures do not contain adequate guidance for SFAS recovery actions. Tracking Number: 26.0647 i
Description:
Plant operating procedures at the time of the 12/26/85 transient did not provide the operators with sufficient guidance for recovering from an SFAS actuation. Resolution: Revise Casualty Procedure C.41, " Recovery from SFAS Actuation" to include detailed guidance on recovery actions. O 4C.9-1 <
1 Closure Requirement: Approval sheet for SFAS recovery procedure. Completion Date: Prior to start of last operator training cycle (in cold S/0). 4C.9.3 Problem 3 BLPBs labeling needs to be revised. Tracking Number: 26.0648
== Description:== The BLPBs for SFAS components that require multiple operator actions to obtain manual control were not labeled with sufficient detail to aid the operators in performing these actions. Resolution: Relabel the applicable BLPBs. Closure Requirement: Work performed by Nuclear Operations. Inspected by System Engineer as complete. Completion Date: Complete. 4C.9.4 Problem 4 Operators need to be trained on SFAS recovery procedures. Tracking Number: 26.0649
== Description:== Operators need to be trained on Casualty Procedure C.41
" Recovery from SFAS Actuation" and Operating Procedure A.70 " Safety Features Actuation System."
Resolution: Train the plant operators on C.41 and A.70. Closure Requirement: Memo from Training Department stating completion of operator training. Completion Date: Prior to initial plant heatup. 4C.9.5 Problem 5 SFAS circuit board / module problems. Tracking Number: 26.0650
== Description:== B&W identified potential problems with cracked module connectors and with circuit board retainer clips which may allow a circuit board to loosen during a seismic event. l l l 4C.9-2 l l
f Resolution: Perform an inspection of the SFAS modules and circuit a boards as detalled in the B&W letters and
) repair / replace as necessary.
Closure Requirement: Completion of HR: 113838 116864 113839 116865 113840 116866 116867 Completion Date: Prior to functional testing of SFAS in cold S/D. 4C.9.6 Problem 6 SFAS SPs contain inaccuracles. Tracking Number: 26.0141
Description:
The SFAS SPs contain typos, references to incorrect procedure steps and other inaccuracles. Resolution: Revise any inaccuracles found in the SFAS SPs. Closure Requirement: Memo from the System Engineer documenting the review of the following SPs and attached approval sheet of SPs requiring revision: SP.200.13, "SFAS Surveillance Calibration" SP.200.09, " Monthly SFAS Surveillance Procedure"
- SP.203.01 A/B, "SFAS Digital Channel 1A/B Refueling
- 1 -
Test" . SP.204.01 A/B, " Refueling interval Reactor Building
- Spray System Loop A/B SFAS Surveillance Test" Completion Date
- Prior to functional testing of the SFAS in cold S/D.
4C.9.7 Problem 7 MUT outlet valve control problems. Tracking Number: 26.0142
Description:
There is some question as to what pushbuttons must be depressed to obtain manual control of the Makeup Tank (MUT) outlet valve following an SFAS actuation. Resolution: Perform a special test to determine what actions are necessary to obtain manual control. O 4C.9-3
Closure Requirement: Completion of WR 111423 and approval sheet of STP 783. Completion Date: Prior to functional testing of SFAS in cold S/D. 4C.9.8 Problem 8 Review of open SFAS Work Request (prior to 07/30/86). Tracking Number: 26.0143
Description:
The Plant Maintenance Daily Work Listing (dated 07/30/86) was reviewed for open SFAS work items that must be completed prior to start up. Resolution: Complete work prior to restart. Closure Requirement: Closure of the following hrs: 111734 115516 116864 111866 116730 116865 114062 116732 116866 114074 116733 116867 114823 116850 111418 114827 116851 113577 114828 116852 108877 114829 116961 111423 114830 116962 114839 114835 116963 112232 114836 113838 114843 113839 114844 113840 Completion Date: Prior to functional testing of SFAS in cold S/D. 4C.9.9 Problem 9 RCS pressure transmitter experiences voltage spikes. Tracking Number: 26.0184
Description:
Shielded instrument cable for PT-21099 (used for the ; interlock on DH Drop Line Valve HV-20002) was l inadvertently run in a power cable tray. As a result, voltage was induced in the signal circuit resulting in spurious actuations of HV-20002 when it was required to be open. Resolution: Reroute new instrument cable in the proper cable tray and reterminate. Closure Requirement: Completion HR# 16925 and close NCR S-5263 and ECN R-0459. 4C.9-4
Completion Date: Prior to initial plant heatup.
-~
s,I s 4C.9.10 I Problem 10 Review of NRC IE Notice 85-94. Tracking Number: 27.0062
Description:
NRC IE Notice 85-94 " Potential for loss of minimum flow paths leading to ECCS pump damage" describes a potential loss of minimum pump flow paths leading to ECCS pump damage following an SFAS actuation. Resolution: Review NRC IE Notice 85-94 for applicability to Rancho Seco. Closure Requirement: System Engineer Report CCL 86.0388. Completion Date: Prior to reactor start up. 1 4C.9-5
... _-. . - _ _ _ . --... - _ . .. --. _ . - ....-.. . . _--- ---- - . - --___ - - ..._. - -__ - . ~
4C.10 Reactor Protection System (RPS) O b 4C.10.1 Problem 1 Potential loss of ground in RPS cabinets. Tracking Number: 26.0286
Description:
A B&W Preliminary Safety Concern identified the concern where an RPS channel may experience a loss of ground to its instrument common without the loss of ground being evident. Given this condition, a single postulated failure in one channel can leave the RPS in an unanalyzed condition. Resolution: Revise appropriate procedures to include testing which verifies that a proper instrument ground exists in the RPS cabinets. Closure Requirement: Approval sheet of SP 200.08 revision and I.108A (B, C, and D). Completion Date: Prior to reactor start up. 4C.10.2 Problem 2 The daily instrumentation surveillance procedure needs to be revised to comply with Tech Specs. and the USAR. Tracking Number: 26.0191
Description:
Surveillance Procedure SP 200.01 " Instrumentation Surveillance Performed Each Shift" does not adequately verify all the requirements of Table 4.1-1 and USAR Section 7.1.2.3.4. Resolution: Revise SP 200.01 to ensure verification of Tech Spec and USAR requirements. l l Closure Requirement: Approval sheet of SP 200.01 revision. l Completion Date: Prior to reactor start up. [ ( 4C.10.3 Problem 3 Electrical terminations in the RPS cabinets need rework. l Tracking Number: 26.0168 i 4C.10-1
== Description:== Inspections of all electrical terminations in the RPS I cabinets were conducted and NCRs written on any ) deficiencies found. l Resolution: Repair / rework all electrical termination deficiencies identified during the inspection of the RPS cabinets and documented by NCRs. Closure Requirement: Closure of the following: NCR-55384 NCR-S5390, ECN-R0615, NCR-55395, ECN-R0636, , NCR-S5396, l NCR-S5403, ECN-R0644 l NCR-S5404, ECN-R0642 l NCR-S5406, NCR-S5408, ECN-R0663, NCR-S5446, ECN-R0614 l NCR-S$429, NCR-55749, 1 NCR-S5787, i Completion Date: Prior to initial plant heat up. 4C.10.4 Problem 4 RPS circuit board retainer clip / module connector problems. Tracking Number: 26.0996
== Description:== Cracking was suspected in the CKT board corners and retainer clips. Resolution: Perform or verify performance of inspection recommended in PSC 3-84 and PSC 10-82. Closure Requirement: Perform or verify performance of inspection recommended in PSC 3-84 and PSC 10-82. l Completion Date: Prior to initial plant heat up. 4C.10.5 Problem 5 Review of open RPS Work Requests (prior to August 13, 1986) required for system operability. Tracking Number: 26.0190 1 4C.10-2
Description:
The Plant Maintenance Daily Work Listing (dated n 8/13/86) was reviewed for open work items that must be ( completed prior to start up and were not addressed by any other problems. Resolution: Close all WRs. Closure Requirement: hrs 106873 111501 109237 115956 115963 116003 115975 116004 115977 116874 116881 117145 116882 116883 116884 117325 117543 117544 Completion Date: Prior to initial plant heat up. 4C.10.6 Problem 6 Add Class 1 T-HOT signal to SPDS. Tracking Number: 26.0372
Description:
The current T-HOT signal supplied to the SPDS is not a Class I signal as required by Reg. Guide 1.97. Resolution: Install Class 1 T-HOT Signals to the SPDS from the RPS and revise calibration procedure. Closure Requirement: Closure of ECN-A5672 and approval sheet of Instrument Calibration Procedure I.lli revision. Completion Date: Prior to initial plant heat up. O 4C.10-3
l l 4C.11 Plant Security System (PSS) ( O 4C.11.1
~
Problem 1 Failed Emergency Battery Lighting Units and dark spots around plant areas (below 0.2 foot candles, Safety concerns in NSEB roof areas due to additions of new HVAC Units. Tracking Number: 26.0458
Description:
- 1. Due to recent lighting surveys and the addition of
, new plant structures, the plant protected area Security Lighting System needs to be upgraded in order to meet the requirements of 10CFR 50.73 and the Rancho Seco Security Plan.
- 2. NCR S-5488 has identified that the photocell of various Emergency Battery Lighting units have failed. The Emergency Battery Lighting units are required by section IIIJ of 10CFR 50 Appendix R for safe shutdown capability of the plant.
Resolution: Install new components per ECNs R-910 (Emergency Battery Lighting Units), R-1029 (Additional Security Lighting and Modification), ECN A-5544 (NSEB Roof Lighting). Closure Requirement: Closure of ECN R-910, R-1029, A-5544 and NCR S-5488. Completion Date: Prior to reactor start up. 4C.11.2 Problem 2 Door AU129 does not have position switch monitor. l Tracking Number: 26.0852
Description:
The District has committed to the NRC that all fire area boundary doors will be provided with position switches. I Resolution: Install door position switch / monitor on door AU129. i Closure Requirement: Closure of ECN R-0921. Completion Date: Prior to reactor start up. O 4C.I1-1
4C.11.3 Problem 3 Open priority 1 work requests. Tracking Number: 26.0885
== Description:== There are two priority 1 work requests: to resolve the following problems: Illumination of 0.2 foot-candles is required. Pole A-97 is below the minimum.
- Investigate all vital access doors for proper operation. Identify doors in need of repair. 1 l
Resolution: Close Work Requests. ! Closure Requirement: Closure of Work Requests 107331 and 120496. l Completion Date: Prior to reactor start up. l O O 4C.ll-2
4C.12 Nuclear Instrumentation (NIS) 4C.12.1 Problem 1 Open Work Requests Priority 1. Tracking Number: 26.0623
Description:
Resolve the following problems: Power range "B" signal causing nuisance alarm. Intermediate range "B" signal causing nuisance alarm.
- Perform I.ll3 on incore thermocouples.
Resolution: Repair the nuisance alarms and perform the calibration surveillance. Closure Requirement: Completion of Work Requests 115146, 122971, and 119880. Completion Date: Prior to reactor start up. 4C.12.2 Problem 2 Open Work Requests Priority 1. O V Tracking Number: 26.0876
Description:
Perform STP.251 at 40% reactor power. Resolution: Perform STP.251 when reactor is a 40% power. Closure Requirement: Closure of Work Request 111500. Completion Date: When 40% power level is achieved. O 4C.12-1 1
4C.13 Annunciator System (ANS) O. 4C.13.1 Problem 1 Review / revise annunciator procedure for H2ES-Windows 84,101. . Tracking Number: 22.0400
Description:
The response section of.the annunciator procedure for 125 V DC "E" Bus trouble (H2ES-84) does not identify each feeder breaker and its respective load. Investigation shows-H2ES-101 has the same problem. Resolution: Revise procedure for Windows H2ES-84 and 101 to include load list and the appropriate operator response for loss of respective loads. Closure Requirement: Revision of Annunciator procedures. Completion Date: Prior to reactor start up. t 4C.13.2 . Problem 2 Open ECNs priority 1. Tracking Number: 26.0515
Description:
ECN R-0871 - Provide IDADS trouble alarm. . A. Delete control room annunciator H2HC-16 " Elevator Trouble."
- 8. Establish H2HC-16 as "IDADS Trouble."
C. Safety to ensure there is sufficient information in the elevators for calling assistance in case of emergency. ECN R-0688 - Provide alarm for pressurizer relief actuation. A. Combine annunciator window H2PSA-57 " Enable Borate" and H2PSA-58 " Enable De-Borate" into one window H2PSA-57 " Enable Borate /De-Borate." B. Establish H2PSA-58 as " Pressurizer Relief Flow." This window will be actuated by retransmission of existing IL/ IDS alarms via MUX-4. Resolution: Provide IDADS trouble alarm and provide alarm for , pressurizer relief actuation. O Closure Requirement: ECN R-0871, R-0688, and completion of HR 114417, 118803, 115290, 117269, and 117268. U Completion Date: Prior to reactor start up. 4C.13-1
-~ _.
4C.13.3 Problem 3 Open work requests. Tracking Number: 26.0514
Description:
While I&C was performing loop calibration on steam generator level (Operate range Channel 8), the annunciator window H2PSB-14 stayed in solid all the time although the audible alarm would make and break. Resolution: Repair annunciator window H2PS8-14 Closure Requirement: Completion of WR 117371. Completion Date: Complete. O l O 4C.13-2
4C.14 Plant Communication System (CSP) 4C.14.1 l Problem 1 Third phone for Control Room. Tracking Number: 15.0182
Description:
A third phone with a separate number is to be provided in the Control Room. This phone is to be used during off-normal conditions. Resolution: Complete installation of third phone. Closure Requirement: Complete HR 112233. Completion Date: Prior to leaving cold shut down. 4C.14.2 Problem 2 Review of Open Priority 1 Work Requests. Tracking Number: 26.0625
Description:
Open Work Requests which should be completed. HR 117602 -- Repair the speaker outside the Diesel Fire Pump Room. HR 105301 -- Repair speakers H7T221, H7J1552 and H9H20. HR 103795 - Repair speaker in the pump alley of -20 Auxiliary building. HR 112464 -- Change crankcase oil and oil filter on microwave emergency diesel generator. Last lab analysis indicated fuel dilution. 1 Resolution: Complete work on Work Requests. Closure Requirement: Work Requests 117602, 105301, 103795, and 112464. Completion Date: Prior to leaving cold shutdown. 4C.14.3 Problem 3 Emergency Communications Alternatives Task Tracking Number: 26.0624 O V 4C.14-1
== Description:== In the event cf a communications room fire, all in-plant and off-site communications will be lost with the exception of sound powered phones and limited plant coverage provided by 5 watt, Channel 2 UHF radios. Plant communications are to be improved by addition of new ROLM node. Resolution: Change system configuration for relocation of UHF radio repeater. Install new ROLM node 2. Closure Requirement: ECN R-1300 Completion Date: Prior to reactor start up. 4C.14.4 Problem 4 Antenna system in the Tank Farm and Pipe Chaseway areas. Tracking Number: 15.0187
== Description:== Unreliable communication may be addressed by new equipment. Resolution: Verify radio communication with accident response equipment operating. Closure Requirement: Test results. Completion Date: Start up. O 4C.14-2
4C.15 Plant Computer System (PCS) t 4C.15.1 Problem 1 Safety Parameter Display System Deficiencies Tracking Number: 26.0854
Description:
The NRC Audit Plan requires upgrade of the SPDS reliability. SPDS displays do not always provide accurate correlation with other Control Room indicators. Due to improper calibration of once through stea:n generator trend recorders and Safety Parameter Display System steam generator level display, the operator is presented with conflicting information. This can lead to over-filling of the steam generator. The SPDS manual does not correctly describe the temperature calibration algorithm. Resolution: Complete Task 753 and install SPDS upgrades to improve SPOS reliability. Closure Requirement: Closure of A-5249, A-5672, R-0952 and Completion of B&W Task 753 Completion Date: Prior to reactor start up. 4C.15.2 Problem 2 The SPDS display may be much less rollable than meters. Tracking Number: 26.0526
Description:
The District's SPDS display of Category 1 Reg Guide 1.97 variables, is one of the first applications of a Class I computer at a Nuclear Plant. Resolution: Perform an investigation on the need for comparison l between the SPOS Reg. Guide 1.97 display and a hypothetical hardwired (typical) display. Closure Requirement: Copy of Engineering investigation report. Completion Date: Prior to reactor start up. 1 lO l 4C.15-1
4C.15.3 Problem 3 Some Radiation Monitoring Points are missing from SPDS. Tracking Number: 26.0529
Description:
The District took credit for Radiation points on DRMS. However, since the DRMS has not been verified and validated (V&V), the points called out in NUREG 0737 should be put on the SPDS. ENC R-1041 adds Main Steam Line monitor points R-15047 and R-15048 to the SPDS. Resolution: Install Building Effluert, Main Steam Line and Radwaste Area Stack Radiation alerts to the SPDS. Closure Requirement: Closure of ECN R-1041 l Completion Date: Prior to reactor start up. 4C.15.4 Problem 4 The NRC has requested that the District compare the SPDS sof tware to Reg Guide 1.152. Tracking Number: 26.0530
Description:
Although there is no requirement for SPDS software to meet the guidance in Reg Guide 1.152, the NRC has asked that we perform a comparison. Resolution: Perform evaluation of SPDS software to meet requirements of Reg Guide 1.152. Closure Requirement: Copy of B&W report. Completion Date: Prior to reactor start up. 4C.15.5 l Problem 5 SPDS is not covered by Tech Specs. l Tracking Number: 26.0531
Description:
SPDS will display the post-accident, Reg Guide 1.97 Category 1 variables. The Tech Specs must be revised. Resolution: Revise Tech Specs. Closure Requirement: Revision of Tech Specs. i Completion Date: Prior to reactor start up. 4C.15-2
c 1 _ .,1 4C.15.6 Problem 6 The SPOS reliability analysis is based on the previous ;,e configuration. , ,' V Tracking Number: 26.0532 !
Description:
The Reliability Study for the SPDS submittal was based on the previous configuration. Since SPDS is being upgraded for Reg. Guide 1.97, the B&W reliability study should be revised. Resolution: Revise SPDS reliability calculation. ; Closure Requirement: Revision of SPDS reliability calculation. . Completion Date: Prior to reactor start up. W 4C.15.7 Problem 7 The Heat-up/Cooldown rate is not displayed in the Control Room. 4 Tracking Number: 26.0855
Description:
During the review of the 12/26/85 trip sequence a concern was raised the the operators in the Control Room did not have the time to manually calculate the
~-
cooldown rate of the RCS during the transient. Engineering has developed modifications to provide two
- dedicated incore trend indications on HlRI. A dedicated, two pen trend recorder will provide an
, average of Class 1 incores and the rate of change of ,
t these Class 1 incores. Lines on the faceplate of the trend recorder will indicate critical temperature rates. l Resolution: Install heat-up/cooldown rate monitoring. Closure Requirement: Closure of ECN-1028. ! Completion Date: Prior to reactor start up. 4C.15.8 i Problem 8 IDADS Annunciator Procedure discrepancies. Tracking Number: 22.0352 1 4
- 4C.15-3 l
- R
ll
,0es'cription: IDADS point (C1904) is not included in the IDADS
..-- annunciator procedure.
,/
This IDADS point (Lil54) is not included in the IDADS annunciator procedure. Annunciator Procedure H2SFA Window 10 setpoints are confusing. Two cases of " reactor trip" and "no reactor
, trip" run together.
Resolution: Update C.39 and Annunciator Procedure H2SFA to correct p
/
the noted discrepancies. Closure Requirement: Approval sheet of C.39 and Annunciator Procedure H2SFA revision. Completion Date: Prior to reactor start up. 1 iI i s
/
O 4C.15-4 l
4C.16 Nuclear Service Coolins Water (NSCW) l . 4C.16.1 Problem 1 Open Work Requests Tracking Number: 26.0208
Description:
The following open Work Requests must be completed.
- 1. WR 116796 -- Due to several recent overheatings, test condition of motor for NSCW Pump P-482A.
- 2. :HR 117254 -- Remove, inspect and repair, if required, ammeter in H2SFA for P-482B in accordance with NCR S-5645.
- 3. WR 104488 -- Change ammeter caution label from 280 amps to 308 amps to reflect change in procedures and process standards.
- 4. WR 107906 -- Replace transmitter and. resistors for level indicator for NSCW Surge Tank "A".
1
- 5. WR 107907 -- Replace transmitter and resistors for level indicator for NSCW Surge Tank "B".
- 6. WR 112709 -- Repair water leaking through weld at NSW-016, inlet to WU Pump LO Cooler.
- 7. WR 1128054 -- Repair leak at flange-body connection for Safety Features valve, inlet to RB Cooler A-500A.
- 8. WR 116540 -- Test / rebuild snubber ISH-50053-2A.
- 9. WR 116406 -- Test / rebuild snubber ISW-50050-6A.
- 10. WR 116535 -- Test / rebuild snubber 19W-050060-3A.
,' 11. WR 116539 -- Test / rebuild n M ar iW-50050-2A.
- 12. WR 115720 -- Perform hydro test o. NSCW Loop "B" (10 year inservice testing).
- 13. WR 115891 -- Calibrate FT-500038 (flow through RB i Eme"Jency Cooler) A-500C.
- 14. WR 116006 -- Replace FT-50001A (flow through RB Emergency Cooler) A-500A.
- 15. WR 109843 -- Calibrate TI-48703 (Decay Heat Bearing O Temperature P261A.
4C.16-1
Resolution: Complete all priority 1 work requests prior to start i up. Closure Requirement: Close work requests. i Completion Date: Initial plant heat up. O O 4C.16-2
l l 4C.17 Nuclear Service Raw Water (NSRW) O V 4C.17.1 i Problem 1 Possible flaking of interior epoxy coating of raw water i piping. I Tracking Number: 26.0144
Description:
The tube side of the NSCH heat exchanger and the steel l elbows in the raw water piping are epoxy coated. Flow in the system could be severely restricted if flaking of epoxy coating occurs. Since the system is the ultimate plant heat sink and provides decay heat and emergency generator cooling, such a failure could be ! quite serious. ' Resolution: Inspect the following equipment:
- 1. P-261A/B 01l Cooler
- 2. E-237B
- 3. A-5298, A-529C, and A-529E
- 4. P-291 A/B 011 Cooler Closure Requirement: HR 117977, HR 117978, HR 120423, HR 120424. HR 120425, HR 120426, HR 120427, HR 120428, NRC-S 4969 Completion Date: Initial plant heat up.
4C.17.2 Problem 2 Open work requests. Tracking Number: 26.0145
Description:
The following is a list of prioritized open work j requests: l
- 1. HR 114619 -- Repair hypochlorite injection pump, P-471A.
- 2. HR 110755 -- Repair leak in piping at NRH-013.
Makeup from diesel fire pump at NSCH heat exchanger "A" in tank farm.
- 3. HR 116232 -- Repair hypochlorite pump P-4718. i
- 4. HR 115719 -- Perform hydro test. This is part of inservice testing. ;
- 5. HR 115760, 115762 -- Measure motor current at breaker and at edgewise ammeter in control room for l
NSRH pump P-472A. (NCR S-4565) : 4C.17-1
- 6. HR 115761, 115764 -- Heasure motor current at breaker and at edgewise ammeter in control room for NSRH pump P-4728. (NCR S-4565) '
- 7. HR 115333 -- Replace relief valve, PSV-43705, on inlet to diesel generator cooling water heat exchanger for generator "A".
- 8. HR 115334 -- Replace relief valve, PSV-47306, on inlet to diesel generator cooling water heat exchanger for generator "B".
- 9. HR 115864 Repair inoperative ammeter for P-472A.
Resolution: Complete all priority one work requests prior to restart. Closure Requirement: See individual work request numbers. Completion Date: Initial plant heat up. 4C.17.3 Problem 3 Review of priority 1 ECNs. Tracking Number: 26.0315
== Description:== ECN R-0328 Replace damaged 5 kV shielded power cable to NSRH pump P-4728. New cable is class lE. The original cable was damaged due to small junction box, improper pull box and conduit. These items will also be repaired. Resolution: Complete all priority 1 ECNs prior to restart. Closure Requirement: ECN R-0328 Completion Date: Initial plant heat up. 4C.17.4 Problem 4 Review of priority 1 NCRs. Tracking Number: 26.0317
== Description:== NCR S-5645 addressed damage to ammeters on panel H2SFA. Ammeter for P-472A was damaged and needs to be inspected and replaced if necessary prior to startup. Damage occurred during replacement of lugs. O 4C.17-2
i Resolution: Inspection, repair ammeter and close NCR prior to restart. Closure Requirement: NCR S-5645 Completion Date: Initial plant heat up. 4 l l l l J i l i i 9 d i i i l t . i i l l l l 4 l 0 i i l l 4C.17-3
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4C.18 4160VAC O 4C.18.1 Problem 1 Work Request review. Tracking Number: 26.0390
Description:
The following is a list of open Work Requests that will be closed before restart:
- 1. WR 111418 -- Troubleshoot Annunciator H2SFA.
- 2. WR 115535 - Remove / replace bus links for Bus D inspection.
- 3. WR 117612 - Troubleshoot DG Breaker (4A08).
Charging springs failure to discharge in disconnect position. Resolution: Complete work requests prior to restart. Closure Requirement: WR 111418, 115535, 117612 Completion Date: Prior to plant heat up. 4C.18.2 Problem 2 Procedure A.58 needs revision. Tracking Number: 26.0460
Description:
Operating Procedure A.58, 4.16KV Electrical System, needs to be revised to resolve various concerns.
- 1. Provide adequate guidance for operation of the GE Switchgear (54A2 and S4B2) installed in the NSEB.
4
- 2. Include load breaker (4E12) Misc. Water Tritium Evaporator in Section 4.18.3.6.
- 3. Include a list of DC control breakers which supply the buses.
- 4. Correctly identify Section 4.19 for Bus S4E2.
Resolution: Revise A.58 and provide training. Closure Requirement: Procedure A.58 and completed training. Completion Date: Prior to plant heat up. O 4C.lG-1
40.18.3 Problem 3 Problem with breaker not closing. Tracking Number: 21.0078.0
Description:
4A2 and 482 bus breaker racking interlocks may prevent breaker closing due to a sticking racking arm when breaker racking tool is removed after the racking in process. NCR S-5700 identifies Breaker 48207 as being racked in, out of alignment when the rackout arm of the breaker failed to engage the rackout mechanism in the cubicle. After repairs were completed, the breaker would not charge its springs in the racked in position. Subsequently, under the direction of a GE representative, Electrical Maintenance replaced the breaker's bent racking arm (HR 116731) and adjusted its operating mechanism (HR 116752). Functional testing verified successful operation. Resolution: Update procedure A.58 and inspect all S4A2 and S482 breakers. Closure Requirement: Update procedure A.58 and incorporate resulting from inspections. Completion Date: Prior to plant loss of power test (LOSP). 4C.18.4 l Problem 4 Surveillance Test revision. Tracking Number: 27.0029
Description:
Surveillance procedures need revision from the l Implementation of Hod 29, ECN A-5013, to include testing of isolation switches required by 10CFR50 Appendix R. Resolution: Revised Surveillance Procedures. Closure Requirement: Completed. Completion Date: Complete. 40.18.5 Problem 5 Modifications required to allow DG testing per Technical Specification 4.6.2. 4C.18-2
P Tracking Number: 26.0398
Description:
To comply with Technical Specification 4.6.2.8 d (Emergency Power System Periodic Testing) and Operational Assessment (OA) 82-21, a time delay relay must be installed in the DG breaker control scheme. The revised Technical Specification, paragraph 4.6.2.B allows the interruption of onsite emergency power by manually tripping the DG breaker directly, instead of shutting down the diesel engine. A time delay relay is required to prevent the breaker from automatically reclosing before the nuclear service bus loads are stripped. -In addition, this relay will also reset the DG breaker anti-pump relay. Resolution: Implement ECN A-0415A. Revise SP.206.02A and B and EM.177A and B. Provide operator training. Closure Requirement: ECN A-0415A. Procedures SP206.02A and B, and EM.177A and B. Completion Date: Prior to plant loss of power test. 4C.18.6 . Problem 6 Inadvertent closure of SKV breaker. Tracking Number: 26.0533
Description:
Breakers have been known to inadvertently close
; following re-energization of control power.
! Cases have been documented at Rancho Seco and other facilities where breakers have inadvertently closed. Two possible causes have been identified:
- 1. A breaker mechanical malfunction, or
- 2. A close signal being present at the time of breaker installation and/or control power energization.
This represents a safety concern to operators while energizing control circuits with cubicle doors open. An equipment concern is also significant such as with the diesel generator feed breakers. i Resolution: Revise Procedure A.58 to increase breaker reliability. Closure Requirement: Procedure A.58 Completion Date: Prior to plant heat up. i 4C.18-3 1 i . .~ .-~ _ . _ _.._._ _ ,. , _ ,.._. _ __..,. _ _ _ _ _ . _ _ . _ _ _ . _ _ _ _ _ _ , . . , _ . _ .._ _ . . _ _ , . _ _,_
4C.18.7 Problem 7 Switchyard low voltage operation. Tracking Number: 20.0349
Description:
IE Notice 84-02 pertains to operation at switchyard voltage lower than analyzed. No procedural implementation of Technical Specifications for bus voltage limits exist. The Technical Specification (Section 3.7) requires operator action at 215KV, 217KV, and 219KV. Presently, the plant computer monitors voltage and alarms at 214KV and 218KV (Reference AP.166). No procedure response is dictated in A.54, "220KV System Operating Procedure." Resolution: Modify monitoring of 220KV bus according to Tech. Spec. Revise Procedure A.54 and B.2. Closure Requirement: Procedure A.54 and B.2. Completion Date: Prior to plant heat up. 4C.18.8 Problem 8 Casualty Procedures need to be written. Tracking Number: 26.0589
Description:
4160V Casualty Procedures are not specific enough to ensure proper operator action. Proper operator action is required in order for the plant to remain in post trip window. Among the items insufficiently covered in the Casualty Procedures are the total loss of offsite power, loss of a start up transform (r (independently and together), loss of normal feed to a 4160V bus, total loss of 4160V bus, etc. The procedures should consider:
- 1. A thoroughly thought out action plan.
- 2. Guidance on how to recognize that a loss of power has occurred.
- 3. Recovery steps.
- 4. Damage control.
- 5. Electrical fault isolation plan.
O 4C.18-4
- 6. Quick reference tables on what is still working.
.g I 7. Automatic response by plant.
- 8. Required operator response including Technical ~
Specification requirements / restrictions.
- 9. Loss of control power.
- 10. Loss of cooling on transformers.
- 11. Verify circulating water companion pumps trip during loss of a 4E bus to prevent runout of companion pump.
Resolution: Write Casualty Procedures as necessary to assure proper operator action. Closure Requirement: Casualty Procedures C.1, C.101, C.102, C.103, C.104, C.105, C.106, C.107, C.108. Completion Date: Prior to plant heat up. 4C.18.9 Problem 9 Open NCRs. Tracking Number: 26.0590
Description:
The following NCRs must be closed prior to restart:
- 1. S-5539 - abrasions and cuts on the 4KV stress cones of Diesel Generator feed cable in Switchgear 48202. (HR 115513)
- 2. S-5499 - insulation damage on 4KV busbar of Switchgear S4A202. ('AR 117043)
- 3. S-5369 - components of 4KV Switchgear S4A202 and S48202 were removed and are not stored in accordance with any procedure; a violation of 10CFR50 Appendix B Criteria VIII.
These NCRs identify hardware concerns which affect Safety Related equipment. Resolution: Close out NCRs prior to restart. Closure Requirement: NCR S-5539, S-5449, S-5369 Completion Date: Prior to plant loss of power test. 4C.18-5
-- ,-,e- - - - , - , - - , - - - , --,------.w- - - . - . --,-e--. . - - - - , - - - - , - - - - - - - ---------,,r-,--- ,-- - - - - - , -,,-- ,, - - -*-
40.18.10 Problem 10 Isolation switches on Train "A" equipment. Tracking Number: 26.0391
== Description:== Installation of isolation switches is necessary to isolate Control Room circuit interface from fire in the Control Room for Class lE equipment in accordance with 10CFR50 Appendix R criteria. The addition of isolation switches, identified in Mod 132 and ECN R-1128 is required to isolate Control Room circuits with the added electrical equipment in the NSEB. Resolution: Add the isolation switches as recomended per ECN R-1128. Revise Procedure EM.144. Implement Training. Closure Requirement: ECN R-ll28. Procedure EM.144, isolation switches per ECN and certification of training accomplished. Completion Date: Prior to plant heat up. 4C.18.11 Problem 11 Open Engineering Change Notice (ECN) review. Tracking Number: 26.0392
== Description:== The following is a list of open ECNs which must be closed prior to restart:
- 1. A-5564 - Final connection to diesel generators GEA2 and GEB2 to the 4160V buses. (Mod 36)
- 2. A-3660Z - Final configuration of the electrical distribution system and controls. (Mod 40)
- 3. R-Oll2 - Reca11brate Class I 4160V bus undervoltage protection trip relays to new engineering tolerance requirements.
The ECNs A-3660Z and E-5564 will be implemented upon completion of testing on the new diesel generators and use of the temporary test circuit. ECN R-Oll2 - The required work has been completed and the existing calibration of relays verified to be within the new engineering criteria. O 4C.18-6
Resolution: Modify electrical distribution system per ECNs. Revise 3 operating and surveillance procedures. Closure Requirement: ECN A-5564, A-3660Z, R-0112. Procedures SP206.02A and B, SP206.07A and B; SP206.08A and B. , Completion Date: Prior to loss of power test. 4C.18.12 Problem 12 Open NCR review. Tracking Number: 26.0415
Description:
The following open NCRs must be closed prior to restart:
- 1. S-5444 - Incorrect selection of Amp butt splice for Class IE SKV cable shield.
- 2. S-5478 - Model Number S4A Bus Alarm relay incorrectly marked on shipping container.
- 3. S-5545 - S482 0/V relay test switch incorrectly terminated. (HR 115583)
- 4. S-5820 - Incorrect crimping tool used on Class lE SKV shield lugs.
These NCRs identify hardware concerns which affect Safety Related equipment. Resolution: Close out NCRs prior to restart. Closure Requirement: NCR S-5444, S-5478, S-5545, S-5820. Completion Date: Prior to loss of power test.
- 4C.18.13 Problem 13 Modification of OV/UV protection.
Tracking Number: 26.0389 l
Description:
Overvoltage occurrence in 4160V AC Class 1 buses, and initiation of the diesel generator unnecessarily challenges the safety function of the system. l Resolution: Modify Relays per ECN R-1045 and revise SP.206.07A and B, and SP.206.08A and B. O l 4C.18-7
l l I Closure Requirement: ECN R-1045. Procedure SP206.07A and B, and SP 206.08A and B. Completion Date: Prior to loss of power test. 4C.18.14 Problem 14 Breaker Lockout / Auto close circuit problem. Tracking Number: 26.0705
== Description:== If an auto close signal exists on a 480V, 4160V, or 6900V switchgear breaker, the breaker will lock itself out during:
- 1. The racking in process,
- 2. A manual trip operation, and
- 3. The resetting of the 486 Aux 111ary trip lockout relay .
After racking in the switchgear breaker and re-energizing the control signal, the 52Y coil will energize and seal in before the springs are charged. In this condition, (i.e., the breaker open and 52Y coil energized), the breaker is locked out. The breaker can also become locked out if a trip signal is initiated from the Control Room while an auto close signal is present. The breaker will trip and then trip free ending up in a lockout mode. When resetting the 486 relay after a breaker overload condition, the same situation will exist. There is no abnormal indication to the operator that a lockout condition exists. The 52Y coil will remain energized until the auto close signal is removed or the DC control power is de-energized. Resolution: Evaluate Procedures and provide tralning. Perform an evaluation and implement approved recommendations. Closure Requirement: Not Available. Completion Date: Prior to plant heat up. 4C.18.15 Problem 15 Loss of offsite power with diesel generator paralleled. Tracking Number: 22.0055 4C.18-8
Description:
If the diesel generator is paralleled with offsite power and a loss of offsite power occurs, the diesel O generator would try to carry all the site non-vital loads supplied from offsite power.
~~
During this condition, the diesel generator feed breaker to the nuclear service bus could trip on overload. This would initiate the unloading scheme and trip open the offsite power feeds from the nuclear service bus. The diesel would not close in until the overload lockout relay (486) is reset and the DC control power is cycled by the operator at the local 4160V breaker. Resolution: Investigste the consequences and provide recommendations. Closure Requirement: 1. Study completed.
- 2. Recommendations implemented.
Completion Date: Prior to heat up. 4C.18.18 Problem 18 DVR requirements for Mods 36 and 40. Tracking Number: 26.0394
Description:
Design Verification Reports (DVRs) for Mod 36 (ECN A-33748 new diesel generator facilities) and Mod 40 (ECN A-3660 final electrical distribution system configuration) have been prepared. DVRs have been required since Revision "0" of Nuclear Engineering Procedure NEP-4109, " Rancho Seco Configuration Control Procedure." ECN A-3748 was prepared 06/27/84 before the DVR requirements were established. Major ECN A-3660 Rev. 4 was also prepared before the DVR requirements. Revision 5 of Major ECN A-3660 was issued to incorporate Sub ECN A-3660Z. NEP-4109 Section 3.4, item 3 states, "DVRs are not required for Major ECNs but are required for each sub ECN." The DVR has been issued for ECN A-3660Z. Resolution: This item la closed. Closure Requirement: N/A Completion Date: Completed. O 4C.18-9
4C.19 Lube 011 System (LOS)
! Problem: 4C.19.1 Priority 1 open LOS work requests (11/4/86)
J Tracking Number: 26.0913
Description:
HR 104813 - Lube oli pump motor hot to touch (P-807) HR 100611 - Check the bearing Lube 011 Relief Valve setpoint. Record the "as found" set point. Set valve to 20 2 2 psig. HR 074964 - Change setting of PSL-86801, 86802, 86803 and 86804 from 145 plus/minus 2.5 to 180 plus/minus 2.5 psig. HR 112890 - Inlet pump V-826 is vibrating badly when oil flow is increased. Vibration can only be stopped by stopping oil flow through centrifuge. HR 104664 - Differential pressure across the filter F-872A is greater than 3 psi. HR 104665 - Differential pressure across the filter F-3728 is greater than 3 psi. HR 107402 - Clean, inspect and rebuild PCV-80302C, if necessary. HR 107398 and HR 116495 - Clean and inspect centrifuge (Y-823). HR 117090 - Cuno Filter (F-8718) crud trap on the bottom of the filter leaks oil around the threads. HR 123014 - Filter (F08738) outlet flange drips oil. HR 123015 - Lube oil leaks from a weld joint on the lube oil cooler (F-800A). HR 114911 - Inlet and outtet flange Ieak 01I on F-873C. HR 114914 - Outlet flange leaks oil on F-873D. HR 123049 - Multiple oil leaks (centrifuge, heater 3). liesolution: Perform work required by work requests. Closure Requirement: HR 104813, 100611, 74964, 112890, 104664, 104665, 107402, 107398, 116495, 117090, 123014, 123015, 114911, 114914, 123049 Completion Date: Prior to condenser vacuum. O 4C.19-1 l 1 L
4C.20 Purification and Letdown (PLS) O 4C.20.1 Problem 1 Inadvertent boron dilution of Reactor Coolant System (RCS). Tracking Number: 20.0075
Description:
If demineralized water is not controlled during partial draindown of the reactor coolant systems, boron dilution can occur resulting in Technical Specification violations or in an inadvertent criticality. The resolution for Operational Assessment 80-003 required that the Domineralized Reactor Coolant Storage Tank (CRCST) Pumps, P-622A & B, be placed under clearance. This has been added to Operating Procedure A.1, Reactor Coolant System. The inlet valve to the DRCST from the deborating demineralizers (RHS-347) was also to be placed under clearance. This was done on a one-time basis, but was not put in procedures. Resolution: Revise Operations Procedure A.1, Sections 6.1, 7.1 and 7.2 to place Shift Supervisor's clearance en valve RWS-347. Closure Requirement: Operations Procedure A.1 Revision. Completion Date: Prior to unit turnover to Dispatch. 4C.20.2 Problem 2 Testing of SFV-22009 Accumulator Tank.
- Tracking Number
- 22.0169 4
Description:
The accumulator tank for SFV-22009, Letdown Line RB Isolation, should be function tested. j Resolution: Test the accumulator tank for SFV-22009. Add testing to a surveillance procedure. l Closure Requirement: SP.203.03 Revision. i
- Completion Date
- Prior to the start of hot shutdown testing.
lO 1 4C.20-1
4C.20.3 Problem 3 Testing of SFV-224013 Accumulator Tank. Tracking Number: 22.0170
Description:
The accumulator tank for SFV-24013, Seal Peturn Reactor Butiding Isolation, should be function tested. Resolution: Test the accumulator tank for SFV-24013. Add testing to a surveillance procedure. Closure Requirement: SP.203.03 Revision. Completion Date: Prior to the start of hot shutdown testing. 4C.20.4 Problem 4 Check valves not in surveillance procedures. l Tracking Number: 20.0134
Description:
Certain check valves have not been included in the surveillance program for testing as required by ASME Section XI, as stated in Memo No. EQC 83-296. The five valves listed in the memo are SIM-078, SIM-079, SIM-081, DHS-059, and PLS-045. Resolution: Revise surveillance procedures to include these valves. Closure Requirement: SP.203.03 Revision SP.203.06 C/D Revision SP.203.02A SP.203.02B SP.203.02C SP.214.01 Completion Date: Prior to completion of ILRT. 4C.20.5 Problem 5 Open work requests. Tracking Number: 26.0311
Description:
The following work requests were determined to require closure prior to startup: HR Number Description 113594 Repair remote control for HV-22001 4C.20-2
108221 Repair body to bonnet leak on HV-22008 112747 Repair packing leak on SFV-22005 (A) V 107843 112099 Rework body to bonnet joint on SFV-22006 Replace 1Imlt swltch on SFV-22006 111517 Perform surveillance test of SFV-22006 112748 Repair packing leak on SFV-22023 115584 Replace motor connection on SFV-23508 111423 Circuit testing of SFV-23508 111518 Perform surveillance test of SFV-23508 112745 Repair packing leak on SFV-24004 Collect IE Bulleting 85-03 Data: 115547 SFV-22005 115548 SFV-22006 115569 SFV-22007 115570 SFV-22008 115571 SFV-22025 115567 HV-23004 115549 SFV-22023 115557 SFV-22004 Miscellaneous Work: 115432 Remove / replace snubber ISW-22000--2A 115443 Test /repatr/ retest snubber ISW-22000-2A 109119 Repair boron analyzer AE-22202 109860 Re-energize letdown flow indication (, following NNI-X work Restore computer point F015 (Letdown
, 112881 Flow High Alarm) to computer monitor 119661 Remove spool piece from nitrogen supply line to make-up tank.
112585 Remove / reinstall blank flange for PLS draining. . Resolution: Complete and close out priority 1 work requests. I Closure Requirement: Closed work requests listed in description. Completion Date: Prior to system functional testing except for 110771 which must be complete prior to criticality. 4C.20.6 Problem 6 Carbon steel valve bolting. Tracking Number: 26.0314
Description:
Letdown valves SFV-22005 and SFV-22006 have had a history of corroding studs / bolts. 4C.20-3 1 l
These valves, as well as SIM-020 and SIM-022, have had three incidents each of body to bonnet leakage. ECN 1-2921, Rev. I was written to allow replacement of carbon steel studs and nuts with stainless ones on Velan valves. Resolution: Replace the studs and nuts on these valves with stainless steel studs and nuts. Closure Requirement: Closed work requests /ECN. Completion Date: Prior to system functional testing. 4C.20.7 Problem 7 Open Nonconformance Reports (NCRs) Tracking Number: 26.0375
Description:
The following NCRs were determined to require closure prior to startup. 1 S-5599 - Motor spilce on SFV-23508-L S-5621 - Valve PLS-030 blocked S-5739 - Cracked power cable to SFV-22025-L Resolution: Complete and close out priority 1 NCRs before start up. O Closure Requirement: NCR's S-5599 S-5621 S-5739 Completion Date: Prior to system functional testing. 4C.20.8 Problem 8 Open Abnormal Tags Tracking Number: 26.0377
Description:
The following Abnormal Tags were determined to require closure prior to startup. 3536 - SFV-22006 torque switch setting 5005 - Temporary piping from DRCST to RHUTs Resolution: Remove and close out priority 1 abnormal tags before start up. O 4C.20-4
Closure Requirement: Removal of abnormal. tags 3536 and 5005 9 Completion Date: Abnormal Tag 3536 - Prior to system functional testing. Abnormal Tag 5005 - Prior to unit turnover to Dispatcher. l f i i 1 i l l l l l l 4C.20-5
l 4C.21 Decay Heat System (DNS) O 4C.21 Problem 1 Elbow on DNS "8" drain lIne has pin hole leaks. Tracking Number: 21.00398
Description:
Visual observation of the P-2618 drain piping revealed leakage from an elbow. Resolutlon: Remove existing drain lIne and instalI additional Iines and isolation valves on DH Pump "A" and "B" prior to start up (ECN R-0498). Review system procedures (valve line-up sheets, etc.) and preventive maintenance requirements to ensure changes are incorporated. Closure Requirement: ECN R-0498 complete. Procedure and PM reviews complete and revisions made as required. Completion Date: Prior to reactor start up. 4C.21.2 Problem 2 Large number of open work requests on DNS. Tracking Number: 26.0069
Description:
A large number of work requests are open on the OHS and should be resolved during the scheduled DHS outages. Some work items are individually required to be performed for compliance with Technical Specifications or to minimize operator action outside the control room. After completion of DHS outages, the following lists will need to be reviewed again for effect on restart. The following Work Requests are required for restart: (As of July 1, 1986) Work Request No Description 85576 Body to Bonnet Leak on HV-26105 99000 OHS-022 has stripped stem. 100299 DHS-003 leaking cover gasket HV-20002 operations to verify no O 106975 packing leakage 4C.21-1
107026 Test PSV-26102 per SP 214.02 107919 Calibrate gage per SP 203.11 107923 HV-20001 operations to verify no packing leakage 108562 HV-20003-L position indication removed 109123 Packing leak on SFV-26006 109766 Unable to operate HV-20002-1 with manual handle 110112 DHS-016 Packing Leak 110270 DHS-485 leak through 110589 SFV-26005 Packing Leak 110633 FE-26004 Leak 110806/(110877) Install mod per ECN 1-5772 111507 HV-26106 Not tested per SP 203.06B 111508 HV-26008 Not tested per SP 203.068 112035 Replace P261A drain piping 112036 Replace P2618 drain piping 112443 Packing Leak on DHS-50 114149 Packing Leak on SFV-26006 114195 Readjust limit switch on HV-26037-L 114196 Handwheel turns on HV-26038-L when stroked electrically 114198 Noisey motor on SFV-25004-M 114200 Borated Crystals on HV-26106 114238 HV-26046 leaks by 114239 HV-26047 leaks by 114467 Packing leak on HV-26106 114604 Scaffolding for HV-25004 inspection 114609 OHS-564 has body to bonnet leak 114653 SFV-25003-L banged and popped when opened 114983 P261A Inboard seal leaks 115030 Root valve leaks by on PI-26103 115344 Inspect DHS-001 115437 Rebuild snubbers 115453 Rebu11d snubbers 115456 Rebuild snubbers 115509 Install pulsating damper on flow element 26048 115554 Collect valve data on HV-26007-L 115555 Collect valve data on HV-26105-L 115556 Collect valve data on HV-26106-L 115558 Collect valve data on HV-25003-L 115566 Collect valve data on HV-25004-L 115568 Collect valve data on HV-26008-L 115717 Perform Hydro 115718 Perform Hydro 115722 Perform leak test Resolution: Close open Work Requests. Closure Requirement: Closed open Work Requests Completion Date: Reactor start up. 4C.21-2
4C.21.3 (' Problem 3 No procedure to perform leak test of HV-20001 and HV-20002. Tracking Number: 26.0071
Description:
No procedure presently exits to perform leak tests of Decay Heat Isolation Valves, HV-20001 and HV-20002. Resolution: Write and perform a procedure to accomplish the required leak testing. Closure Requirement: Procedure written, approved, and successfully performed. Completion Date: Reactor start up. 4C.21.4 Problem 4 LLRT not performed on containment penetration No.29. Tracking Number: 26.0072
Description:
Containment Penetration No. 29 is not required to be LLRT'ed per Technical Specification 4.4.1.2 however, it falls under 10 CFR 50 App J, Part II.H.3: " ..(valves) required to operate intermittently under post accident ( conditions." Valves HV-20001, HV-20002, and HV-20003 are required to operate for post accident long term cooling per USAR Section 14.2.2.5.2. Resolution: Review of LLRT program by Licensing with respect to 10 CFR 50 App J to determine testing requirements. Write procedure and perform LLRT if deemed necessary by Licensing. Closure Requirement: LLRT program review by Licensing complete. If necessary, procedure written, approved, and successfully performed. Completion Date: Reactor start up. 4C.21.5 Problem 5 Failure of DHS Suction Line Valve with a S/G Tube Leak. Tracking Number: 26.0079
Description:
A steam generator (s/g) tube leak at Oconee II resulted in a reactor shutdown and subsequent delayed opening of a decay heat drop line valve. The apparent cause of O the decay heat drop line failure at Oconee was a bent 4C.21-3
valve stem. This system had been replaced three (3) times before due to overtorquing by the motor operator. Resolution: Revise emergency operating procedures to incorporate Babcock and Wilcox (B&W) Technical Base document (Task 195 dated 04/04/85) for once-Through Steam Generator (OTSG) Tube Rupture and Coincident unavaliability of DHR. Review Decay Heat Removal (DHR) suction valves (HV-20001 and HV-20002) for correct torque swltch settings and proper valve and operator match. Closure Requirement: Procedure revisions and MOVATS program on valves HV-20001 and HV-20002 complete. Completion Date: Initial plant heat up. 4C.21.6 Problem 6 Failure of Gate Valves to open due to thermal or hydraulle binding. Tracktng. Number: 26.0158
Description:
There have been several instances in the Nuclear Field of safety-related, motor-operated gate valves falling to open due to thermal or hydraulle binding. Resolution: Investigate occurrences from other plants and the history of similar failures at Rancho Seco. Identify valves which have failed and perform survelilance and/or special tests to verify current operability. Train operations personnel in the phenomenon and possible corrective actions (i.e. heating or cooling l valve). l Closure Requirement: Testing requirements identified and tests performed as necessary. Supplemental operator training completed. i Completion Date: Initial plant heat up. l 4C.21.7 Problem 7 Full Flow tests are not being performed on check valves DHS-015 and DHS-016. 4C.21-4
Tracking Number: 26.0404
Description:
Full flow testing of valves DHS-015 and DHS-016 must be (V9 performed during cold shutdowns lasting more than 72 hours if not performed in the past three (3) months. Resolution: Revise surveillance procedures (s) and perform the required testing. Closure Requirement: Procedure (s) revised and testing successfully performed. Completion Date: Reactor start up. 4C. 21.8 Problem 8 NPSH calculations for the Decay Heat Removal Pumps do not cover all potential modes of operation. Tracking Number: 26.0405
Description:
Existing NPSH calculations for the Decay Heat Removal Pump suction from the Reactor Building Sump assume single train operation. However, under certain conditions, the operator may have both trains in operation. Resolution: Nuclear Engineering to review the NPSH calculations for adequacy and prepare new calculations as necessary to verify the NPSH for all potential modes of operation. Closure Requirement: Calculation review complete and new/ revised calculations issued as required. Completion Date: Reactor start up. 4C.21.9 Problem 9 DHS Abnormal Tags Tracking Number: 26.0406
Description:
Tags indicate abnormal system conditions. Tag No. Condition 3517 Torque switch setting on HV-26105 has been changed. 5009 Temporary clamps and seals installed on drain line for Pump-2618 to stop leakage. Resolution: Clear Abnormal Tags prior to restart. O 4C.21-5
Closure Requirement: Clear Abnormal Tags Completion Date: Initial plant heat up. 4C.21.10 Problem 10 Reactor Building Emergency Susp Design Verification Tracking Number: 26.0486
== Description:== A review of the B&W Guide Specification IAS-5 and the present sump configuration indicate there are several differences. The sump performance has not been verified by testing. Also, changes have been made to inlet piping. Testing has been determined to be undesirable. As an alternative, an engineering review should be performed to verify adequacy of the Reactor Building Emergency Sump. Resolution: Perform a design verification of the Reactor Building Emergency Sump. Closure Requirement: Complete design verification. Completion Date: Reactor start up. 4C.21.11 Problem 11 Excessive DHS discharge pressure. Tracking Number: 26.0487
== Description:== Under worst case conditions, operation of the OHS pumps at flows of 3,000 GPM or less may result in the pump discharge pressure exceeding the 450 psig design pressure of the discharge piping and the DHS coolers (E-260A&B). Resolution: Equipment and piping design will be upgraded as applicable to meet or exceed the maximum pressures which will be experienced by the system during operation. Closure Requirement: Close NCR S-5973 Complete upgrade of system design pressure. Complete procedure revisions as necessary. Completion Date: Initial heat up. O 4C.21-6
-a _
. . . . - ~ . _ . - - . . . .. . . -
4' , . n
' a ',
\c 4C.22 Reactor Coolant System (RCS) '
/ ,
p I C'4C.22.1 L /,_ y)
'g l Problem 1 Confusion over cooldown rate limits. C' s
i Tracking Number: 15.0132 1
, 1 1
l i
Description:
There has been confusion over what constituted 100 ' degree per hour cooldown rate. (i.e. is it 1.67 ! t degrees per minute or no more than 100 degrees in any one-hour period?). There is a Tech Spec interpretation based on guidance from B&W that defines 100 degrees per hour as 1.67 degrees per minute (below 550*F) with ' allowance for a maximum of 15 degrees deviation at any t
/ '
time in the cooldown. Resolution: Revise plant operating procedures (82 ai B4) to reft.kct , i the current BIN interpretation of 100 degree per hour. " Train operators on revised procedures. s Closure Requirement: Completed. i Completion Date: Complete. i 4 4C.22.2 j Problem 2 EMOV block valve concerns. , Tracking Number: 26.0890 i
Description:
- 1. The EMOV block valve is usually closed during ',
j normal operation; the EMOV is therefore not ! available to prevent challenges to the code safety ;, i valves. '
- 2. The block valve position should be vertfled open J when operating the EMOV.
- 3. Circuit breaker (2B182) is labeled "PORV Block l
Valve HV-21505," not with the name of the valve as known, "EMOV Block Valve HV-21505." a Resolution: 1. Revise operating procedures to require the block j valve be normally open during operation, t
- 2. Research whether specific instructions to check the i block valve position in the E0P steps described
- above should be made, laplement changes as ,
i necessary. l 3. Revise name on breaker 25182 to read "EMOV Block ] Valve HV-21505." : i Closure Requirement: Revised E.04, E.06, E.07, 8.2. ' Completion Date: Prior to initial heat up (HTUP). : 4C.22-1
,h r
4C.22.3 i ' Problem 3 Analyze conditions of 12/26/85 transient. ( Track f rig' Number: 26.0888
Description:
- 1. Determine minimum level reached by the pressurizer s ,
and potential for reactor vessel head void.
- 2. B&W to analyze potential for core lift.
+
Resolution: 1. Provide calculation to show that the pressurizer emptied.
- 2. Provide document to show that no damage to the fuel assembly hold down springs will occur after possible core lift.
Closure Requirement: Completed per;
- 1. B&W calculation i
- 2. B&W Doc. 51-115:655-00 e
Completion Date: Complete. 4C.22.4 o Problem 4 Rosemount RTD Transmitters (EQ).
! Tracking Number: 26.0898
Description:
Extend qualifted life of RTD transmitters tag numbers TE-21029, TE-21030, and TE-21034 to Cycle 8 outage. Rosemount RTD transmitters TE-21029, TE-21033, and
, TE-21034, which are located inside the Reactor Building, have a qualtfled life of 10 years based on an EQ Aging Analysis performed in SMUD calculation Z-EQP-E0146. The qua11 fled life of these RTDs effectively expires May 19, 1987, based on the aging l analysis performed by EQ. Due to the current outage which started in December 1985, the replacement date of May 19, 1987, is no longer valid. Investigation by the EQ group has determined that since December 1985, the l Hot Leg RTDs have been constantly exposed to low
! temperatures, thereby reducing the thermal stress to which the platinum elements of the RTDs are exposed. Thus, the replacement of these RTDs will be extended to Cycle 8 outage, which is scheduled for January 19, 1989. Resolution: Revise the environmental qualification file Z-EQP-E0146, and MARSS file ERPT-E0077 to extend the
- qualifled Iife to a replacement date of January 1989 during Cycle 8 outage.
Closure Requirement: Revised EQ flies Z-EQP-E0146 and MARSS flie ERPT-E0077. Completion Date: Prior to Reactor Startup (RXSU). 4C.22-2
I f 4C.22.5 Problem 5 RCP power relays have been determined unrollable. Tracking Number: 20.0546 , p'
Description:
During calibration of RCP power relays, four were found l to be defective. One phase balance relay, and three under power relays. ' Existing W11 mar Electronics Model l 21-172, phase balance relays and Model 21-171, ! underpower relays are a constant problem concerning l calibration, component overheating and failures. I Received concurrence from B&W to disconnect the phase balance relays and replace the underrower relays with
. ITE under current relays.
Resolution: , Remove phase balance relays and replace underpower ! s' relays with ITE undercurrent relays. Perform EM.144, l " Testing of Protective and Control Relays," per ECN R-0899. l Closure Requirement: ECN R-0899, NCR S-5557 and NCR S-5598. l Completion Date: Prior to initial plant heat up (HTUP). 4C.22.6 Problem 6 Clarify emergency operating procedure for excessive heat transfer. Tracking Number: 26.0887
Description:
Ouring the 12/26/85 transient, operators had difficulty reconciling the dichotomy between avoiding the pressurized thermal shock (PTS) region and regaining minimum pressurizer level (according to NUREG 1195). Resolution: Revise emergency procedure E.05 and the Rules to de-emphasize pressurizer level when adequate subcooled mergin is maintained, and ensure the rule for HPl throttling takes precedence over usintaining a minis.us pressurizar level. Closure Requirement: Revised E.05. Completion Date: Prior to reactor start up (RXSU). l 4C.22.7 Prob ha 7 Thermal shock of the reactor vessel analysis, Tracking Number: 20.0274 f 4C.22-3
Description:
Generic letter 81-19 deals with the probability of thermal shock to the reactor vessel due to overcooling and injection of cold water into the RCS. SAW-1791 addressed PTS for B&W plants, but SMUD was not included in the analysis. Resolution: Complete analysis prior to restart. Closure Requirement: B&W analysis. Completion Date: Prior to initial heat up (HTUP). 4C.22.8 I Problem 8 Casualty Procedure C.11 " Pressurizer System Failure" requires revision. Tracking Number: 26.0403
Description:
C.11 does not address the following:
- a. Loss of two pressurizer level instruments from a single cause. C.11 states that if two instruments
. agree, use them,
- b. Malfunction of the pressure control system.
Resolution: Revise Casualty Procedure C.11 as follows:
- a. Provide an exception to alert operator if signal conversion cabinet "A" falls,
- b. Include loss of pressurizer heaters and failure of spray valves, including multiple failures such as both spray and block valves failed open.
Closure Requirement: Revised C.ll. Completion Date: Prior to reactor start up (RXSU). 4C.22.9 Problem 9 High Point Vents - Control and Indication. Tracking Number: 26.0402
Description:
- 1) Operating procedure A.1 (RCS) and A.3 (Pressurizer and PRT) do not give the operator guidance on how to energize the high point vents for operation. At the present time, controls for powering the vent valves are de-energized for Appendix R reasons and a procedure in the E0P's tells the operator to use these valves but not how to energize them.
4C.22-4
- 2) High point vents are installed on the A&B hot legs and the pressurizer for removal of noncondensible I
(A V l gases in accordance with the emergency procedures. Operating procedure B.1 " Plant Critical Checks" requires surveillance procedure SP214.03, " Locked Valve List" to be completed prior to going critical. HV20533 and HV20579, RCS A-Loop high point vents, are closed and power removed per i SP214.03. This is done in compliance with 10CFR50 Appendix R. This action removes valve position indication at panel H2SP. NUREG 0737 requires valve control and indication in the control room. Resolution: Investigate the Appendix "R" requirements and revise procedures A.1 and A.3 as appropriate (instructions for eaergizing controls or leaving controls energized). Closure Requirement: Revised A.1 and A.3. i Completion Date: Prior to reactor start up (RXSU). I 4C.22.10 Problem 10 Open NCRs. Tracking Number: 26.0401 s l
Description:
Open NCRs were reviewed to determine which required closure prior to restart. Resolution: Close the following NCRs prior to restart: S-5263 Reroute instrumont cabIe to XV-20002 due to induced voltage spikes which cause the valve to close. S-5354 Replace damaged Once Through Steam Generator (OTSG) Manway Studs. S-5359 Reset torque switches on Iimitorque operators (sample isolation valves). S-5729 Replace damaged power operated relief valve (PORV) block motor. S-5864 Improper weld on pressurizer relief valve-discharge pipe. Closure Requirement: Close the above NCR's. Completion Date: Prior to initial heat up (HTUP). 4C.22.11 Problem 11 Open ECNs. Tracking Number: 26.0400 4C.22-5
Description:
Open ECNs were reviewed to determine which required closure prior to restart. Resolution: Complete the following Priority 1 ECNs prior to restart: R-0688 Instali Pressurizer Rellef Flow Annunciator. A-5772 Install Temperature Measurement device on Decay Heat Line (Temporary - to obtain data for hot leg level temperature compensation). Closure Requirement: R-0688, A-5772. Completion Date: Prior to initial heat up (HTUP). 4C.22.12 Problem 12 Administrative lodine limits. Tracking Number: 20.0401
Description:
During a Steam Generator Tube Rupture (SGTR), primary coolant enters the secondary coolant. Our 08/11/85 response to generic letter 85-12 states that SMUD has implemented an administrative limit of 0.8 micro curies /cc for lodine in the primary system. This limit is not reflected in plant procedures. > Resolution: Limit should be included in SP.202.01 " Reactor Coolant Chemistry" and in AP.306 " Chemistry and Radiochemistry Manual," Section Ill. Action statements are to be developed in case this administrative limit is exceeded, and placed in Casualty Procedure C.7 "High Activity in Reactor Coolant." l l Closure Requirement: Revised SP.202.01 Revised AP.306, section III Revised Casualty Procedure C.7 Completion Date: Prior to reactor start up (RXSU). 4C.22.13 Problem 13 Provide guidance for handling loss of cooling accidents (LOCAs) during heatup and cooldown. Tracking Number: 20.0011
Description:
There is inadequate procedural guidance to ensure core cooling in the event that a LOCA occurs when SFAS has been bypassed (i.e., during heatup or cooldown). O 4C.22-6
Resolution: Revise Casualty Procedure C.3 to direct operators to m appropriate emergency procedures in this event. Revise I appropriate procedures to cover this event, as necessary. Closure Requirement: Revised Casualty Procedure C.3. 1 Completion Date: Prior to reactor start up (RXSU). 4C.22.14 Problem 14 Open Work Requests. Tracking Number: 26.0162
Description:
A review was made of open Work Requests, as of 07/01/86. The following Work Requests were identified as requiring closure prior to restart: Snubber testing: 113203 116408 116417 115408 116409 116418 115439 116411 116420 116501 116412 116421 116520 116414 116422
- Oh l 116521 116551 116415 116416 116423 116424 l
Required testing: 109273 Test gauge for SP.18 115414 Bench Test PSV-21507 115415 Bench test PSV-21507 115715 RCS Hydrotest (H.S.D.) Instrument calibration: 108798 PT-21092 MOV Inspections: 112205 HV-21505-L 112206 HV-21514-L 112207 HV-21516-L 112208 HV-21514-L 112209 HV-21505-L 112212 HV-21517-L 116149 HV-21505-L 116261 HV-21505-L Coolant 110775 Tygon level indicator Containment 110012 vent bottles and tubing Boundary: 110013 vent bottles and tubing 110798 Manways and handholes 110799 Manways and handholes 110800 Manways and handholes 110844 Manways and handholes Q 110846 Manways and handholes 112085 Manways and handholes 114472 RCS-031 blind flange 4C.22-7
Valve repairs: 104550 RCS-017 seat / packing leaks 99379 HV-21516 seat leakage 108248 HV-21516 relap seat 108843 HV-21514 seat leakage 69591 RCS-026 seat leakage 69592 RCS-028 seat leakage 69593 RCS-032 seat leakage 69594 RCS-034 seat leakage Resolution: Close the identified WRs prior to restart. Closure Requirements: Closed WR. Completion Date: Prior to initial plant heat up (HTUP). 4C.22.15 Problem 15 Pressurizer relief valve acoustic monitors are unreliable. Tracking Number: 21 0263.A
Description:
Acoustic monitors of pressurizer safety valve leak detection system indicate valve leakage in cold shutdown. Resolutlon: Calibrate acoustic monitors. Closure Requirement: Calibrate Per I.036 " Pressurizer Relief Valve Position Monitor Functional Check and Calibration." l Completion Date: Prior to initial plant heat up (HTUP). l 4C.22.16 Problem 16 Possible Stress Corrosion Cracking (SCC) of reactor vessel anchor bolts. l Tracking Number: 20.0007
Description:
Preliminary Safety Concern 9-81 dealt with the possible susceptibility to stress corrosion cracking of the reactor vessel support skirt anchor bolts. When anchor bolts with a high yield strength ( 120 ksi) are used and the bolts are preloaded to greater than 707. of ultimate strength, there may be an SCC concern. Resolution: Verify that the SMUD RV anchor bolts are A193 Grade B7, with a minimus yield strength of 105 ksi. O 4C.22-8
Closure Requirement: Write a letter to B&W addressing this issue and closing PSC 9-81. Completion Date: Prior to initial plant heat up (HTUP). 4C.22.17 Problem 17 Pressurizer relief valve setpoint drift. Tracking Number: 20.0003
Description:
Following reactor trip #69, the code safety valve on the pressurizer lifted at 2360 psi, which is 140 psi less then the setpoint. The setpoint drift is believed to have been caused by setting the relief valve before the valve fully warmed up to operating temperature. l Resolution: Insert temperature requirements into MT.003 ! (Pressurizer Code Relief Pressure Verification) to ensure that relief valves are at operating temperature when the setpoint is tested. The temperature requirement should be based on steady state temperature data taken on Work Request 77794. Closure Requirement: Revised MT.003. l Completion Date: Fill and Vent of RCS. 4C.22.18 Problem 18 Inadequate supports on pressurizer retlef valve discharge piping. Tracking Number: 19.0015
Description:
Pressurizer relief valve discharge piping was not originally analyzed for blowdown reactions. Blowdown I forces could cause ductile failure involving distortion of the line which would block discharge flow. Pipe supports are to be modified based on reanalysis. Resolution: Modify the supports. Completion of ECN A-4615 will close this problem. Closure Requirement: ECN A-4615. Completion Date: Initial plant heat up (HTUP). O 4C.22-9
4C.22.19 Problem 19 Training on phenomena of an empty subcooled RCS pressurizer. Tracking Number: 15.0064
Description:
During the 12/26/85 transient the pressurizer emptied, but the RCS remained subcooled because HPI flow began to exceed the rate of contraction shortly after the pressurizer emptied. A possible Reactor Vessel (RV) head bubble may have helped to maintain subcooling. Resolution: Provide training on this phenomena. Closure Requirement: Completed. Completion Date: Complete. O O l 4C.22-10 r.
4C.23 6900VAC A 4C.23.1 Problem 1 Open work requests. Tracking Number: 26.0351
Description:
The following is a list of open work requests that require closure prior to re. start:
- 1. WR# 104152 - Perform bus 6A C.T. and P.T. test per EM 188-189. Wire check cubicle positions per EM 192.
- 2. WR# 104153 - Perform bus 6B C.T. and P.T. tests per EM 188-189. Wire check cubicle positions per EM 192.
Investigation: This testing has been completed. Resolution: Complete work prior to restart. Closure Requirement: Close WRs. Completion Date: 11/05/86 4C.23.2 Problem 2 Procedure needed to verify torquing on bus connections. Tracking Number: 26.0352
Description:
Switchgear faults can develop when bus connections are not torqued to manufacturer's specification. Electrical Maintenance develop a procedure to verify torqued connections. Investigation: A report from Crystal River Unit 3 indicated loose torquing bus connections which resulted in a phase-to-phase fault on their 6.9 kV system. During power ascension testing at Palo Verde, a 13.8 kV bus experienced a phase-to-phase fault. The investigation revealed inadequate torquing of bus connections. Resolution: Procedure shall be implemented verifying bus connections are torqued to manufacturer's specifications. Closure Requirement: Close WR# 120026 and 120027. Completion Date: Loss of off-site power test. 4C.23-1
4C.24 Main Feedwater System (MFW) 4C.24.1 Problem 1 Training on MFW Flow Recorders and Auxiliary Steam. Tracking Number: 15.0129
Description:
Due to overcooling events that have occurred, it is necessary to have operator training on the MFH total flow recorders and the importance of auxiliary steam on the MFH system. Resolution: Conduct MFW training. Closure Requirement: Submit closure document to QA. Completion Date: Prior to leaving cold shutdown. 4C.24.2 Problem 2 Overcooling of the RCS as a result of operator inability to control feed flow from the control Room. Tracking Number: 26.0734 s
Description:
Operator actions to manually reposition main feed control valves after a loss of main feed panel control power established a demand signal on the inoperative controllar which resulted in a feedwater transient, and subsequent overcooling of the affected RCS loop, when control power was restored. Resolution: Revise emergency procedure E.05, Excessive Heat Transfer, steps .2 and .3. Closure Requirement: Adoption of new procedures. Completion Date: Prior to heat up. 4C.24.3 Problem 3 Procedural deficiencies may cause overfilling of the OTSGs and possibly damage main steam lines. Tracking Number: 16.0002.A O 4C.24-1
Description:
The NRC has expressed concern regarding the capability of plant procedures to prevent overfilling the OTSGs and flooding of the main steam lines. Resolution: Review MFW procedures A.50 and E.02 and revise as necessary. Closure Requirement: Adoption of new procedures. Completion Date: Prior to heat up. 4C.24.4 Problem 4 Difficulties in transferring feedwater control from AFW to MFW. Tracking Number: 16.0003.A
Description:
During plant trip #73, problems occurred when switching feedwater control from Auxiliary Feedwater (AFH) to MFH occurred. 1 Resciution: Revise procedure A.51, " Auxiliary Feedwater System," Section 7.9 to allow a more coordinated transfer from AFW to MFW control. Closure Requirement: Adoption and new procedures. Completion Date: Prior to heat up. l 4C.24.5 Problem 5 C:susity procedure C-10 does not require feed system valve operability verification. Tracking Number: 20.0041
Description:
Failure to verify operability and position of MFH System start-up and control valves contributed to an overcooling event at San Onofre, Unit 2. A similar situation could occur at Rancho Seco. Resolution: Revise casualty procedure C-10 to direct the operator to verify main feed start-up and control valve operability. Closure Requirement: Adoption of new procedures. Completion Date: Prior to heat up. l 0 4C.24-2
4C.24.6 Problem 6 MFW Restart - required System work requests. Tracking Number: 26.0156
Description:
WR Number Description Resolution 78028 FWS-109 Attempted to replace Replace gland or valve. packing gland - could not get bonnet off. 99547 FWS-013 'A' MFH loop check valve. Modify valve to allow use of new hinge pin to fit new bushing - work starter (A-5661). 100095 FV-20575 Valves leak by Disassemble, inspect, and 100096 FV-20576 excessively and repair as necessary - 100097 FV-20525 have not been stroke test per SP.214.01. 100098 FV-20526 inspected for a long time. 105526 Main Steam Stop to K307 turbine Repair as necessary. steam chest has excessive leakage. 107707 P317A HFP inboard seal leaks. Record shaft clearances. Replace inboard / outboard seals - rework and realign Pump. 110913 FWS-015 'A' HFH loop iso. Vlv. Inspect valve intervals. Previous inspection revealed dry grease in operator and broken capscrew. Closure Requirement: Complete WRs 78028, 99547, 100095, 100096, 100097, 100098, 105526, 107707, and 110913. Completion Date: Prior to OTSG Secondary Hydro. 4C.24.7 Problem 7 MFW System ECNs - completion recommended prior to start up. Tracking Number: 26.0357 O 4C.24-3
Description:
HR Number Description Resolution A-5415AB Install motor operators on Install the motor operators FHS-015 & 016. as part of EFIC, QTS # 26.0134 ECN A-5415. R-0699 Replacement of Limitorque Replace internal PVC internal, FV-20529 and insulated wire with FV-25030. qualifled nuciear grade wire; reference IE Notice 86-03 and 10 CFR 50.49. l l R-0894 Expanded use of redesigned Modify MFW valves FWS-001, hinge pin on Anchor / Darling 002, and 031. check valves. Cracks were found in FHS-013, 014 welds. Closure Requirement: Complete ECNs A-5415AB, R-0699, R-0699B, R-0984. l Completion Date: Prior to OTSG Secondary Hydro. 4C.24.8 Problem 8 MFW System NCRs - completion recommended prior to start up. Tracking Number: 26.0358 i
Description:
NCR Number Description Resolution S-5744 TS 4.14 requires engineering Perform the engineering evaluation to determine if any evaluation to the review. safety-related component or system has been affected. This snubber failed to meet lock-ur, velocity criteria when teste6. S-5818 Rear bracket on snubber 6SH-32125 Repair the anchor bolts is loose on the concrete anchors. or relocate the holes. This is attached to the MFH piping just upstream of FV-20575. Closure Requirement: Close out NCR S-5744 and S-5818. ! Completion Date: NCR S-5744 - Prior to leaving cold shutdown. NCR S-5818 - closed 11/18/86. O 4C.24-4
4C.24.9 Problem 9 The " fall as is" air test for MFW flow control valves D)' needs improvement. Tracking Number: 26.0726
Description:
Recommendation #4 of OA 83-6 called for revision of SP.210.04A to verify that the " fall as is" function operated properly for the NFW flow control valves (main and start up). Resolution: Revise SP.210.04A to verify the " failed as is" function. Closure Requirement: Complete revision of SP.210.04A. Completion Date: Prior to start OTSG Secondary Hydro. 4C.24.10 Problem 10 Both MFP low discharge pressure alarms occur intermittently. Tracking Number: 26.0729 N
Description:
Intermittent alarms were received from both MFP pressure switches (PSL-31705, 31706) on H2YSB window number 51 and 62. A review of NCR S-5223 and Abnormal , Tags 3897 and 3898 was performed. Discussion with plant personnel indicates the cause of the alarms was due to induced voltages in the interconnecting cables. Resolution: Determine the root cause of the alarms and take the necessary steps to correct the problem. Closure Requirement: Close NCR S-5223 and clear Abnormal Tags 3897 and 3898. Completion Date: Prior to initial plant heat up. J , 4C.24-5 1 l _ _ _ _ _ _ _ _ . _ __ - _ _ - . _ _ ~
4C.25 Main Steam System (MSS) O 4C.25.1 i Problem 1 Uncertain lift and blowdown set point for main steam safety valves. Tracking Number: 26.0319
Description:
MSSVs have opened below the design set points on numerous occasions in the past. Resolution: Adjust set points on all MSSVs IAW accepted standards. Closure Requirement: Complete work request. Completion Date: Prior to Reactor (Rx) start up. 4C.25.2 Problem 2 Damage to flashing on the main steam bypass Iine to condenser indicates possible water hammer. Tracking Number: 15.0244 p
Description:
Examination of the main steam bypass line in the vicinity of the damaged flashing is necessary. Investigation: Damage to insulation on the main steam line to the turbine bypass line indicates a water hammer might have occurred. The most probable cause is an interference which occurred due to thermal growth of the line. l Resolution: Review work requests 100745 and 100746 and all associated documentation. Closure Requirement: P+ID mark up showing system walkdown closes work requests 100745 and 100746. Completion Date: Prior to Rx start up. 4C.25.3 Problem 3 Drifting of the 4A feedwater reIlef valye set point and failure of the pegging steam pressure controller. Tracking Number: 26.0320 4C.25-1
Description:
Set point drift of the 4A shell side relief valves and the simultaneous malfunction of the pegging steam pressure control. Resolutlon: Prepare PM task for PSV-32455, PSV-32456, PIC-32454, and PIC-32453. Closure Requirement: Printouts of new PM tasks. Completion Date: Prior to Rx start up. 4C.25.4 Problem 4 Develop method to calculate releases from MSSV opening. Tracking Number: 26.0322
Description:
Develop some timely method to assess which valve (ADV or MSSV) is open, and the length of time it has been open. Resolution: Write software to facilitate rapid calculation of off-site dose releases using information provided by the acoustic monitors. Closure Requirement: Install software change. Complete job number IDA-0362. Completion Date: Prior to Rx start up. 4C.25.5 Problem 5 Develop procedure to calculate releases from MSSV opening. Tracking Number: 26.0323
Description:
Develop a procedure which will allow Control Room and Health Physics personnel to calculate radiological
- releases based on how long and how many main steam safety relief valves lifted (as described in 26.0322).
Resolution: Revise AP.509 i Closure Requirement: Revised procedure. Completion Date: Prior to Rx start up. f I 4C.25.6 Problem 6 The Main Steam Safety Valves acoustic monitors are unreliable. l 4C.25-2
Tracking Number: 26.0221 N'
Description:
The MSSV acoustic monitors are unreliable. Resolution: Recalibrate alI acoustic monitors using procedure 1.036A and establish testing and maintenance procedure 1.036A on a regular basis. Closure Requirement: PM HR# P65433. Revise task 3954. Completion Date: Prior to Rx start up. 4C.25.7 Problem 7 The temporary Atmospheroic Dump Valve back-up nitrogen system is not adequate and does not fully meet 10CFR50 Appendix R design criteria. Tracking Number: 26.0325
Description:
The temporary Atmospheric Dump Valve back-up nitrogen system is not adequate and does not fully meet 10CFR50 l Appendix R design criteria. Resolution: Install class 1 back-up bottled air for Turbine Bypass Valves and Atmospheric Dump Valves. ( Closure Requirement: ECN R-0859 and ECN A-5743. Completion Date: Prior to MSS functional testing. 4C.25.8 Problem 8 Turbine stop valve failed to close on turbine trip. 1 Tracking Number: 26.0166
Description:
Trip Report #53 states that one of the turbine throttle stop valves failed to close on turbine trip. Instead, j one of the turbine governor valves was closed. Resolution: Perform Surveillance Procedure SP.210.03C during start up. Closure Requirement: Completed Surveillance Procedure SP.210.03C. Completion Date: During power ascension. 4C.25-3
1 l 4C.25.9 Problem 9 Remote ADV loading station is not equipped with i adequate indication. j Tracking Number: 26.0326
Description:
Operators are unable to observe main steam system pressure while operating the ADVs from the remote i location. j Resolution: Move the remote ADV loading station to the remote i shutdown panel. Closure Requirement: Close ECN A-5415-P ; Completion Date: Prior to Rx start up. 4C.25.10 Problem 10 Direct position indication for TBVs and ADVs is not provided. Tracking Number: 26.0327
Description:
Direct position indication of TBVs and ADVs is not provided. Resolution: Install valve position indicator lights in the Control Room.
- Closure Requirement
- ECN package R-0828.
l Completion Date: Prior to Rx start up. 4C.25.11 Problem 11 Modify Operating and Casualty Procedures for TBVs/ADVs to accommodate the new EFIC modifications. l Tracking Number: 26.0329 l
Description:
Operating and Casualty Procedures for TBVs and ADVs l need to be updated based on new EFIC modifications. Resolution: Nuclear Operations to review and rewrite procedures. Closure Requirement: 1) Copies of revised procedures.
- 2) Hemo SRT 86-240.
l Completion Date: Prior to operator training on procedure rewrite. 4C.25-4
4C.25.12 "x
) Problem 12 Unnecessary loads on MSS may exist during plant trips.
Tracking Number: 20.0161
Description:
Procedural guidance on removing unnecessary steam loads from the main steam header following a trip is not adequate. Resolution: Revise B.4 to add guidance reminding operators to remove unnecessary steam loads. Closure Requirement: Revised procedure B.4 Completion Date: Prior to Rx start up. 4C.25.13 Problem 13 Pegging steam pressure control subsystem needs modification. Tracking Number: 23.0030
Description:
System upsets can occur when pegging steam is enabled to the second and fourth point feedwater heaters. O Resolution: Modify pegging steam control hardware via ECN. Closure Requirement: Closed ECN package (to be supplied). Completion Date: Prior to Rx start up. 4C.25.14 Problem 14 Provide operator training for new TBV/ADV controls modification. Tracking Number: 26.0331
Description:
Operator training will be required prior to restart due to extensive modifications to the TBV and ADV controls being made this outage. Resolution: Conduct retraining for operating personnel in the revised Operating, Casualty, and Emergency Procedures. Closure Requirement: Documentation of training held including attendance sheets. Completion Date: Prior to leaving hot shutdown. a 4C.25-5
4C.25.15 Problem 15 Install motor-operated isolation valves on TBVs and ADVs. Tracking Number: 26.0332
Description:
During the December 26, 1985 transient, failure of the ICS caused all of the TBVs and ADVs to go to the 50 percent open position. This contributed to the subsequent plant overcooling. The inability of the operators to isolate the TBVs/ADVs was identified as a contributing factor to the event in the NUREG 1195 report. Resolution: Install two new motor-operated isolation valves on the turbine bypass headers. In4 tall motor operators on two existing main steam valves (MSS-017) and MSS-018). Closure Requirement: ECN A-5415-AC and A-5415-AD. Completion Date: Prior to secondary Hydro (5415-AD). Prior to hot SBy (5415-AC). 4C.25.16 Problem 16 Complete open Work Requests prior to restart. Tracking Number: 26.0335
Description:
A review of the open Work Requests has produced the following items to be completed prior to start up: WR # Description 111348 B-0TSG pressure on IDADS drifting. 110403 Recalibrate B-0TSG pressure transmitter. 115947 Acoustic monitor XY-26571B bad PS. 115827 Broken instrument air line PV-360148. 104766 Steam trap XCV-30805 leakage. l 106614 A-moisture separator reheater 1" leakoff valve bad. 108626 HV-30225C packing leak. 114581 HV-20565 packing leak. 114616 PV-20564 leak on air operator. 115825 PV-20565 air operator leak. 115324 PV-20561 loose handwheel and packing leak. 110090 PV-20561 high stroke time. 105679 HV-20597 packing leak. 104853 MSS-019 packing leak. I 104852 MSS-021 packing leak. I 110479 HV-20570 leaks by. 108451 MSS-020 packing leak. 4C.25-6 t . . _
107815 PV-36014A packing leak. p 107814 MSS-491 packing leak. I 105826 Sample valve MSS-499 packing leak. ( 115356 PM XCV-20616. 115355 PM XCV-20612. 107290 Packing leak HV-20570. 115358 PM XCV-20571. 115357 PM XCV-20617. 115361 PM XCV-32457. 115360 PM XCV-20611. 115363 PM XCV-20615. 115364 PM XCV-20618. 115365 PM XCV-20570. 107888 PM XCV-35070. 112560 Leaking steam sample valve MSS-003. Resolution: Complete outstanding work requests. Closure Requirement: Closed work requests. Completion Date: Rx start up. 4C.25.17 Problem 17 Closure of System NCRs prior to restart. Tracking Number: 26.0336
Description:
The following NCRs require closure prior to restart: NCR # Description NCR S-5328 MSS trap valves in violation of design - criteria 57. NCR S-5576 Turbine stop valve closing time outside of US/.R Chapter 14 specification. NCR S-5512 Various items on TBVs. NCR S-5507 Various items on ADVs. Resolution: Repair for NCR S-5512 and S-5507. Void NCR S-5576. Revise USAR S-5328. Closure Requirement: Completed work requests and closed NCR S-5512. S-5507, and S-5328. Copy of voided NCR S-5576. Completion Date: Rx start up. O 4C.25-7
4C.25.18 Problem 18 Closure of ECNs prior to start up. Tracking Number: 26.0337 .
Description:
The following ECNs require closure prior to start up:
- 1. ECN R-0914 AF Main Stm to AFH Turbine ref. IE Bulletin 85-03.
- 2. ECN R-0914 AG Main Stm to AFH Turbine ref. IE Bulletin 85-03.
- 3. ECN R-0699 Replace Limitorque internal wiring.
- 4. ECN R-0861 Hanual/ automatic control of TBVs.
Resolution: Install hardware per above listed ECNs. Closure Requirement: Closed ECN packages. Completion Date: Rx start up. O l O 4C.25-8
4C.26 Gland Steam & Condensate (GSC) 4C.26.1
' Problem 1 Review of Open GSC work requests (prior to October 3, 1986) that are Priority 1.
Tracking Number: 26.0666
Description:
The following Priority I work requests are open.
- 1. HR 107897 PSV-30125 (Seal Steam Supply to HP Turbine) is to be replaced. Ref. NCR S-5189.
- 2. HR 112467 TV-30319 (Gland Steam Desuperheating Temperature Control Valve).
- 3. HR 107153 Loose A-344A housing drain line connection.
Resolution: Complete work requests. Closure Requirement: Close above work requests. Completion Date: Reactor start up. O V I l 4C.26-1
4C.27.1 Vital 12SVDC 4C.27.1 Replacement of batteries BA, BB, BC, and BD. Problem 1 Tracking Number: 26.0211
Description:
Engineering Change Notice No. ECN-R-0608 was issued for the removal of the existing auxiliary building 125 Volt
, DC Vital Batteries BA, BB, BC, and BD and their associated racks. The ECN.also provides for the procurement and installation of replacement batteries and racks.
Resolution: Replace batteries BA, BB,'BC, and BO.
- a. Removal of original celis and racks.
- b. Installation of new racks and cells.
- c. Performance of two-hour service test.
- d. ' Add electrolyte to battery cells as necessary and establish baseline specific gravity levels.
Closure Requirement: Work Requests: 114808, 114812, 116194 NCR-S-5353 ECN-R-0608 Completion Date: Prior to Loss of Offsite Power Test. 4C.27.2 .I l Problem 2 Modifications to battery bus loads due to upgrading of 120 Volt AC inverter fed buses.
~
Tracking Number: 26.0212
Description:
As a means of upgrading the power supply reliability for the ICS, NNI and other important plant loads, the 120 Volt AC Vital buses in the Auxiliary Building will be supplied from the new vital inverters in the NSEB. l The original Auxiliary Building vital inverters will be disconnected. Resolution: Revise the battery calculation and perform battery load ' tests, if necessary. Closure Requirement: Revised battery calculation; SP.206.04 performed, if necessary. Completion Date: Prior to Loss of Offsite Power Test. 4C.27-1
4C.27.3 Problem 3 Replacement of circuit breakers in NSEB DC distribution panels to satisfy alarm criteria. Tracking Number: 26.0213
Description:
ECN A-4687 was written to install replacement circuit breakers in DC panels SOA2, S082, SOC 2, S002. This change implements the disposition of NCR S-3111 which recognizes the fact the panel breakers installed by the vendor under H00-040 did not conform to the purchase specification nor the USAR. In addition, replacement battery supply circuit breakers are required to conform to the USAR. Resolution: Replace circuit breakers in NSEB DC distribution panels, and new alarm circuits. Closure Requirement: ECN-A4687, NCR-S3111, WRs 105913-105919, 107052-107058. Completion Date: Prior to Loss of Offsite Power Test. 4C.27.4 Problem 4 Addition of battery charger failure alarms. Tracking Number: 26.0214 .
Description:
Battery charger failure alarms will be added to DC buses SOA2, S082, SOC 2, and S002 within the NSEB. This modification will be performed via ECN-A3660Z. This change represents one out of eight modification categories in ECN-A3660Z, which is intended to establish the final configuration of the Electrical Distribution System being reconfigured via M00-040 since 1982. Specifically, this modification adds undervoltage relays and associated IDADS alarms to the four 125 Volt DC Vital buses within the NSEB. Resolution: Install battery charger failure alarms in NSEB DC distribution panels. Closure Requirement: ECN-A3660Z, Item 7. Completion Date: Prior to Loss of Offsite Power Test. 4C.27.5 Problem 5 Investigate battery chargers transient of 06/23/86. Tracking Number: 26.0216 4C.27-2
L I
Description:
The charging of battery BA was being transferred from charger H4BAC to charger H4BA when H4BA tripped on high O' voltage. Special Test Procedure (STP)-960 was performed 08/13/86 in order to duplicate the charger transient. Resolution: Perfora special test and document the results. Closure Requirement: Work Requests 113001, 113002, 116182, and 116189; STP-960 closure. Completion Date: Complete 4C.27.6 Problem 6 Revise the battery maintenance procedures. Tracking Number: 26.0217
Description:
The weekly, monthly, and refueling interval maintenance procedures for the 125 Volt DC Vital batteries need to be revised. Resolution: Complete revisions and approvals to the battery maintenance procedures. Closure Requirement: Revised battery maintenance procedures. Completion Date: Prior to reactor start up. 4C.27.7 Problem 7 Review of open Work Requests Tracking Number: 26.0224
Description:
Various open Work Requests require closure prior to system restart. Resolution: Close open Priority 1 Work Requests. Closure Requirement: WRs 112226, 112278. l Completion Date: Prior to Loss of Offsite Power Test. 4C.27.8 Problem 8 Systems "A" and "B" are not fully independent when using standby battery chargers. i Tracking Number: 26.0257 i 4C.27-3 l
'. f
Description:
ECN-Ril27 will relocate the standby battery charger , feeders to their same "A" or "B" system HCCs. Resolution: Modify the power supplies for the standby battery chargers. . Closure Requirement: ECN-R1127. Completion Date: Prior to Loss of Offsite Power Test. 4C.27.9 Problem 9 Lack of 125 VDC Casualty Procedures. Tracking Number: 26.0984
Description:
Casualty Procedures have not been provided for 125 Volt DC power. - Resolution: Provide Casualty Proceduras. Closure Requirement: New Casualty Procedures. ', Completion Date: Prior to reactor start up. < 4C.27.10 Problem 10 Open ECH review. Tracking Number: 26.0986
Description:
Open ECNs associated with this system were reviewed for Priority 1 criteria. Resolution: Close open ECNs. Closure Requirement: ECN-A3748, ECN-A5415. l ! Completion Date: Prior to reactor start up. 4C.27.11 Non Vital 125VDC Problem 11 Station batteries BE and BF may be near end useful life. Tracking Number: 26.0445
Description:
Two of the original plant Gould batteries have been determined to be near the end of their useful lives based upon deterioration of the positive plates and the rate of cell failure. 4C.27-4 n<
y ?( ,,, 2 e 3
)
Resolution: Replace batteries BE and BF. ; t Closure Requirement: ECN R-0609 and NCR S-5377. Completion Date: Complete. 4C.27.12 Problem 12 Modification to battery bus loads due to upgrading of-120 volt AC Inverter fed buses. I Tracking Number: 26.0446 ;
's
Description:
Four 120 Volt AC vital buses and one non-vital bus in
- the Auxiliary Building will be supplied from the new
,, vital and non-vital inverters in the NSEB. The Auxiliary Building inverters will be disconnected.
Resolution: a. Modify per ECN-R-0927.
- b. Review and revise operating procedures and train operators.
Ciosuie Requirement: ECN-R-0927. CompletionDate: . Start up. 4C.27.13 Problem 13 Review of open Work Requests. Tracking Number: 26.0449
Description:
Various open Work Requests require closure prior to J system restart. The open Work Requests associated with
,' Priority 1 problems as of 09/22/86 are:
, 112278 Verify correct connector hardware.
/ 116168 Replace battery SE intrarack cables and gr hardware.
116169 Replace battery BF hardware. 116172 Provide backup DC during BE, BF rework. 116179 Provide backup DC from BN1 for BB replacement. 117629 Recalibrate voltmeter on distribution panel F.
\t Resolution: Close open Priority 1 Work Requests listed above prior to restart.
Closure Requirement: Closed Work Requests. Completion Date: Prior to start up. y,
- O 4C.27-5
- ?
S -
l-4C.27.14 Problem 14 Lack of Casualty Procedures. Tracking Number: 26.0453
Description:
Casualty Procedures have not been provided for 125 Volt DC cower. Resolution: a) 22.0302 Provide Casualty Procedures for 125 22.0403 VoIt DC non-vital buses. 20.0358,(A) b) 22.0397 Provide tripping air compressor in new Casualty Procedure for bus E. Closure Requirement: Casualty Procedures. Completion Cate: Start up.
\
G 0 4C.27-6
4C.28 480VAC 4C.28.1 Problem 1 Cable routing does not meet 10CFR50 Appendix R regJirements. Tracking Number: 21.0078.F
Description:
Redundant cable trains routed from the Nuclear Service Electrical Building (NSEB) to the Auxiliary Building do not meet the 10CFR50 Appendix R criteria. Resolution: Install a 3-hour fire barrier between the redundant trains. Closure Requirement: ECN R.0803 Completion Date: Complete. 4C.28.2 Problem 2 Rewrite casualty pro'edure C.109 to enhance guidance for limiting loads wlan powering the 3A2 bus from the 3A bus. Tracking Number: 22.0066
Description:
Casualty Procedure C.109 does not caution the operator to limit the loads when re-energizing bus 3A on loss of offsite power. Resolution: a. Add a caution to C.109. Closure Requirement: Revised C.109. Completion Date: Prior to the Loss of Offsite Power / Engineering Safety Features Test. 4C.28.3 Problem 3 Rewrite casualty procedure C.111 to enhance guidance for Iimiting ioads when powering the 382 bus from the 38 bus. Tracking Number: 22.0072 O 4C.28-1
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== Description:== Casualty Procedure C.lli does not caution the operator to limit the loads when re-energizing bus 382 on loss of offsite power. Resolution: Add a caution to C.111. Closure Requirement: Revise C.Ill. Completion Date: Prior to the Loss of Offsite Power / Engineering Safety Features Test. 4C.28.4 Problem 4 Rewrite Operating Procedure A.59. Tracking Number: 22.0062
== Description:== A.59, the 480 Volt Operating Procedure, does not give the operator clear direction for re-energizing buses 3A2 and 382 from their normal source when offsite power is lost and then regained. { Resolution: Give clear directions in C.109 and C.111. Closure Requirement: Revised C.109 and C.lll. Completion Date: Prior to the loss of Offsite Power / Engineering Safety Features Test. 4C.28.5 Probles 5 Rewrite Casualty Procedure C.109. Tracking Number: 22.0064
== Description:== Section 2.7 of Procedure C.109 states that no immediate response is needed for the loss of Diesel Generator "A" room vent and exhaust fans. This section should reference Technical Specification 3.7.2.C (Diesel Operability). Resolution: Revise Casualty Procedure C.109 to reference Technical Specification 3.7.2.C. Closure Requirement: Approval sheet of C.109 Revision 8. Completion Date: Prior to heat up. O 4C.28-2
l f 4C.28.6 ( Problem 6 Rewrite Casualty Procedure C.111. Tracking Number: 22.0252
Description:
Section 2.7 of Procedure C.lil states that no immediate response is needed for the loss of Diesel Generator "B" room vent and exhaust fans. This section should reference Technical Specification 3.7.2C (Diesel Operability). l Resolution: Revise Casualty Procedure C.111 to reference Technical Specification 3.7.2.C. Closure Requirement: Approval sheet of C.lli Revision 8. Completion Date: Prior to heat up. 4C.28.7 Problem 7 Open work request review. Tracking Number: 26.0297
Description:
The following is a list of open work requests that must be closed prior to restart:
- 1. HR 117562 -- Normal closure of MCC cubicle door 2A165 (Diesel Generator 'A' Control Circult) will trip the breaker inside.
- 2. HR 116216 - Breaker S2C264 (Aux Lube Oil Centrifuge Y-823) trips inadvertently.
- 3. HR 117140 -- Pump P-605B (Flash Tu k) will not run unless BLPB ON is held IN. Suspect sea in contact is not closing.
- 4. HR 108951 -- Complete deletions / additions in cubicle 52-3A215 and 52-38215 (feeders to MCC's-2A4 and 2B4) per ECN A-3748 Rev. 3. Reference HR 104030.
- 5. HR 112939 -- Inspect harness wiring on breaker 52-3A202 (feeder to Bus S3A2).
- 6. HR 112935 - Inspect harness wiring on breaker 52-38202 (feeder to Bus S3B2).
O O 4C.28-3
- 7. HR 116997 -- Remove abnormal tag #3462 to S28400 space heaters after temporary circuit is disconnected and permanent power is restored.
Reference NCR S-5601.
- 8. HR 116998 -- Remove abnormal tag #3463 to S2A400 space heaters after temporary circuit is disconnected and permanent power is restored.
Reference NCR S-5601.
- 9. HR 113317 -- Repair security lighting for the month of August, 1986.
Resolution: Complete all work requests prior to restart. Closure Requirement: HR 117562, 116216, 117140, 108951, 112939, 116997, 116998, and 113317. Completion Date: Prior to the Loss of Offsite Power /SFAS test. 4C.28.8 Problem 8 Open ECN Review. Tracking Number: 26.0382
== Description:== This ECN must be closed prior to restart; A-5198 - Form, lug and terminate coiled cables to provide permanent feeders to MCC S2C8 and S207 in DGB. Resolution: Complete ECN prior to restart. Closure Requirement: ECN A-5198 Completion Date: Prior to the Loss of Offsite Power / Engineering Safety Features Test. 4C.28.9 Problem 9 Open Nonconformance Report (NCR) Review Tracking Number: 26.0383
== Description:== The following open NCRs must be closed prior to restart:
- 1. S-5879
- 2. S-4964
- 3. S-5462
- 4. S-5463 Resolution: Complete all open NCRs prior to restart.
4C.28-4
Closure Requirement: NCRs S-5879, S-4964, S-5462 and S-5463. (ONs) Completion Date: Prior to the Loss of Offsite Power / Engineering Safety Features Test. 4C.28.10 Problem 10 Various Drawing Discrepancies. Tracking Number: 26.0384
Description:
The one-line diagrams and elementary drawings are used by the operators to write clearances and operate plant equipment. The operators cannot perform these tasks effectively with incorrect drawings. Resolution: Complete all drawing changes with respect to one-lines and elementaries prior to restart. Closure Requirement: OTS Numbers NCRs 22.0660 S-5748 S-5660 22.0494(C) S-5682 S-5641 22.0239 S-5867 S-5642 22.0498(B) S-5868 S-5718 Completion Date: Prior to heat up. 4C.28.11 Problem 11 Motor Control Center Trouble Alarm when breakers are manually tripped. Tracking Number: 26.0385
Description:
MCCs, equipped with status position indicators, are unable to differentiate indication between a fault and a manually initiated trip. As a result, an MCC Trouble alarm will occur when a breaker is manually tripped (e.g., for clearance purposes). Resolution: Install one bank of disconnect switches (one location, pre-wired) on each of the ten Westinghouse motor control centers. Closure Requirement: ECN R-1157 Completion Date: Prior to heat up. 4C.28-5
4C.28.12 Problem 12 Requirement for isolation switches on Train "A" equipment. Tracking Number: 26.0607
== Description:== It was discovered, during the Appendix R Audit in 1985, that Rancho Seco does not have the minimal amount of "A" train isolation switches. These switches are required to isolate class IE Control Room circuits during a fire in the Control Room. ECN R-1128 will accomplish this task. Resolution: Add the 480 volt isolation switch required by Appendix R to circuit breaker 3A202. Closure Requirement: ECN R-ll288 Completion Date: Prior to heat up. O O 4C.28-6
l 4C.29 Main Generator Seal 011 System (MGS/ SOS) O 4C.29.1
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Problem 1 Open work requests to be completed prior to start up. Tracking Number: 26.0633
Description:
Plant Maintenance activities which should be performed i prior to start up. ; l Resolution: Investigate the cause of the problem in WRs 109180, l 123009, and 123150 and correct the problems. Closure Requirement: Main Generator System I l HR 104132 - Inspect and tighten PMG rotor pole support mounting bolts as described in attached Westinghouse OMM-057. HR 109180 - Investigate temperature element (TE-84556-H2 Cooler Outlet Hot Gas Temperature) which appears to be failed low. hrs 115534 & 115544 - Configure bus to permit backfeed of Main and Unit Auxiliary Transformers per Abnormal Tags 3878, 4444, 4445, 4446, 4447, and 4448. HR 123191 - Replace desiccant in X-98A desiccant filter. SEAL OIL SYSTEM HR 111325 - Fix the oil leak on the DC back-up Seal Oil Pump (P-839). HR 117512 - Replace valve 50S-597 in the air-side seal oil pump suction line. HR 119372 - Fix the oil leak in the air-side seal oil pump (P-838). HR 123009 - Calibrate the high pressure seal oil back-up pressure indicator (PI-80306). HR 123150 - Air / Hydrogen-side pressures should be within 22 inches H20. Air-side pressure is 4.3 inches higher than the H20. Investigate. HR 119077 - Remove PDCV-80307 and inspect, clean (if necessary), then reinstall. O v HR 119078 - Remove PDCV-84501 and inspect, clean (if necessary), then reinstall. 4C.29-1
HR 119079 - Remove PDCV-84502 and inspect, clean (if necessary), then reinstall. HR 119080 - Remove PDCV-83806 and inspect, clean (if necessary), then reinstall. HR 120471 - Remove PDCV-83804 and inspect, clean (if necessary), then reinstall. Completion Date: Prior to start of hot shutdown testing. 40.29.2 Problem 2 Fix seal oil cooler leak and clean up seal oil skid. Tracking Number: 26.0641
Description:
System walkdown revealed oil leakage on the Seal 011 skid. Resolution: Complete oil cleanup and repair leak. Closure Requirement: Complete HR 114584. Completion Date: Prior to start of hot shutdown testing. O i l l O 4C.29-2
l l 4C.30 Once-Through Steam Generator (OTSG) ! O 4C.30.1
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Problem 1 Disagreement exists between OTSG level instrument i channeI indications. Tracking Number: 15.0045A-1
Description:
The Shift Technical Advisor's comments following the December 6, 1985 transient indicated disagreement between the OTSG level as indicated by the Safety i Parameter Display System and the strip chart recorders. Subsequent investigation revealed that this was not the case. Resolution: Revise Action item Closure Report "SPOS vs. Strip Chart for 0TSG Operate Level". Closure Requirement: New revision to supersede the 01-28-86 revision. Completion Date: Prior to start up. 4C.30.2 Problem 2 OTSG shell thermocouples are not reliable. Tracking Number: - 23.0026
Description:
Emergency Operating Procedures for Steam Generator Tube Rupture require the operators to maintain OTSG tube to shell at less than 100 degrees F for a normal SGTR cooldown and 150 degrees F for an emergency SGTR cooldown. However, OTSG shell thermocouples have been unreliable in the past. Resolution: Complete a work request. Repair and calibrate the existing system. Closure Requirement: Complete W/R. Completion Date: Prior to start up. 4C.30.3 Problem 3 The existing 0TSG-level instrumentation is not anyironmentally qualifled and could not be expected to provide accurate indication in a post-LOCA or post-MSLB environment. Tracking Number: 20.0215 4C.30-1
Description:
An elevated temperature in the Reactor Building, resulting from either a LOCA or a MSLB, would cause density changes in the reference legs of the OTSG level instruments, and changes in the transmitter operating characteristics. These changes will result in an inaccurate indication. Resolution: Install environmentally qualified level transmitters, and insulate reference legs associated with those transmitters. Closure Requirement: ECN 1-5415 Completion Date: Prior to start up. 40.30.4 Problem 4 Various steam generator related temperature limits specified in various plant procedures are not compatible. Tracking Number: 26.0235
Description:
AP.151 states that 25 degrees F is the maximum allowable difference between OTSG Downccmer Outlet Temperature and RCS T-cold. However, A.6, B.4 and DP 1101.01 all state that 25 degrees F is the maximum allowable difference between OTSG downcomer inlet temperature and RCS T-cold. A.47 and AP.152 state that 350 degrees F is the maximum allowable difference between final feedwater temperature and OTSG average shell temperature. However, A.6, AP.151, and B.4 state that 350 degrees F is the maximum allowable difference between final feedwater temperature and OTSG Downcomer Outlet temperature. Resolution: Revise the appropriate procedures to reflect the l correct limits and correct perleds of appilcability. Closure Requirement: New revisions to selected procedures issued. Completion Date: Prior to start up. 40.30.5 Problem 5 There are twelve (12) open work requests on the OTSGs which must be closed out prior to start up. l Tracking Number: 26.0241 l 4C.30-2
Description:
The following work requests must be closed out prior to start up:
\ \ WR# Subject 105765 Insulation Replacement 107048 Insulation Replacement i 110792 Insulation Replacement 110793 Insulation Replacement 110846 Insulation Replacement 110798 Manway or Handholes Cover Replacement 110799 Manway or Handholes Cover Replacement 110844 Manway or Handholes Cover Replacement 11085 Manway or Handholes Cover Replacement 112085 Manway or Handholes Cover Replacement 110510 Replacement of Nitrogen Line 110800 Removal of Temporary Ventilation System Resolution: Complete the work requests prior to' start up.
Closure Requirement: The above listed work requests must be closed out. Completion Date: Prior to start up. O 4C.30-3
l 4C.31 Site Resorvoir System (SRS) 4C.31.1 l 1 Problem 1 Review of open work requests. l Tracking Number: 26.0720
Description:
There are seven work requests that are open on the SRS. Resolution: Perform the work as required to complete the work requests. Closure Requirement: HR# 109909, P-431B, screen spray pump trips, required for operations. HR 119048, Y-429A, screen will not turn when energized. Electrical problem exists. HR 119918, PSL-43103, Deadband between set and reset points is excessively high (30 psig). There is also excessive corrosion on the internal components. HR 123274, LI-43507, Reservoir level indicator not giving indication. Please investigate and repair. Tried to reset and all indication died. d HR 101200, SRS-002, valve cannot be manually operated. Suspect lever shaft pin maybe broken. HR 115128, HV-43608A, valve will not close on timer or manually. Inspect / repair as needed. HR 10537, HV-436088, this valve has been secured for one year due to oversight. Valve appears to be unable to fully close. Completion Date: Prior to Steaming Secondary System. J O 4C.31-1
4C.32 Instrument Air System (IAS) d 4C.32.1 Problem 1 Auxiliary Feedwater (AFW) Control Valves fail open upon loss of IAS. Tracking Number: 26.0227
Description:
The actuators for the AFH control valves are spring loaded to fall open on a loss of IAS. With AFH in use after a plant trip, the fall open mode would be detrimental to the stability of the plant and could c.ause a rapid cooldown similar to the 12/26/85 transient. Resolution: Install a component level air bottle backup system. Closure Requirement: ECN R0859 A5415H Proc. A40, C.23 Completion Date: Prior to EFIC function testing of AFH system. 4C.31.2 Problem 2 The failure of the IAS, either a total loss of air or a single component' failure, has the potential to cause a severe transient due to a loss of control capabilities of vital components. Tracking Number: 26.0339 Resolution: Perform a system review of the IAS to identify i potential, unacceptable failure modes. Closure Requirement: QCI-12 closure form Completion Date: Initial plant heat up. 4C.32.3 Problem 3 Letdown fIIter back flush valves malfunction, Isak by, and are a major source of leakage of the IAS. Tracking Number: 26.0340
Description:
Letdown filter valves HV-22001, 22002, 22401, 22501, and 22201 malfunction, leak by the seat, and leak instrument air excessively. Manual, local operation of 4C.32-1
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the valves is required due to the fault actuators, thus, As Low As Reasonably Achievable (ALARA) standards are not maintained. Resolution: The valves and actuators have been replaced ECN 4-0286 is closed. Closure Requirement: QCI-12 closure form Completion Date: Initial plant heat up. 4C.32.4 Problem 4 There is no operational backup system to the IAS/SAS to prevent a loss of air due to component failures or a - loss of off-site power. Tracking Number: 26.0341
== Description:== A loss of IAS during power operation or following a trip could result in a major transient due to the loss of valve and equipment control functions. The lack of any auto-start back-up system requires operators to both start the existing diesel driven air compressor and reposition the valves manually. Also, this compressor does not provide instrument quality air. Resolution: Install an auto-start diesel driven air compressor. Closure Requirement: ECN A-5233 NCR S-6029 Completion Date: Completion of Decay Heat Removal System Outage. 4C.32.5 Problem 5 Plant Operating Procedures do not reflect the modifications being implemented and do not provide adequate guidance for existing lAS components. Tracking Number: 26.0230
Description:
Operations Procedures A.40 and C.23 do not include the new modifications to the system, nor is C.23 a functional procedure regarding corrective operator actions upon loss of IAS. Resolution: Revise Plant Operations Procedures A.40 and C.23 Closure Requirement: Approval of A.40 and C.23 revisions. Completion Date: Prior to completion of all IAS modifications. 4C.32-2
4C.32.6 Problem 6 Main Feedwater valves FV-20575, FV-20576, FV-20525 and N FV-20526 require backup air suppiies for elosure,_per the EFIC modification. Tracking Number: 26.0231
Description:
The subject control valves require a backup air supply to provide a motive force to close the valves upon receipt of an EFIC signal. Resolution: Install the component level backup air supply for the main feedwater valves per ECN R-0859. Closure Requirement: ECN R-0859 ECN 1-5415Y Completion Date: Prior to EFIC function testing on MFH valves. 4C.32.7
- Problem 7 The Atmospheric Dump Valves (ADVs) fall closed upon loss of IAS.
Tracking Number: 26.0232
Description:
The ADVs use air open/close actuators with back-up - accumulators. Upon a loss or decrease of instrument air pressure, the accumulator supplies air to close the valves. Upon a loss of IAS Control Room actuation of the valves is inhibited. During a transient, the ADVs provide a control of the main steam header pressure. At present, upon loss of IAS, the only actuation is local, manual operation of the valves or manual . connection of a nitrogen bottle "six-pack" that does , not have the necessary tubing fittings. In addition, the accumulators do not meet the applicable code. 2 Resolution: Install a component-level back-up air supply to the ADVs per ECN 1-5743. The accumulator tanks will be removed. Closure Requirement: ECN A-5743 A-54152 Completion Date: Prior to EFIC function test on ADV's. l 4C.32.8 ) Problem 8 Review of open work requests on IAS/SAS - Priority 1. l l 1 4C.32-3
Tracking Number: 26.0114
Description:
The following open work requests exist on the IAS and SAS and should be addressed prior to restart: HR# Oescription System Status 106196 Hole in flapper clogged, SAS Open receiver not pressurizing. 106661 PCV-91505 needs to be reset IAS Open/ 111205 Closed 107378 Copper lines damaged to IAS Open/ 114225 CCW isolation valves Open 110091 Air leak from valve in IAS Open distribution box 9C 112593 Replace packing rings on SAS Open aftercooler E-901A 112594 Replace packing rings on SAS Open aftercooler E-901C 114640 Noise on inboard end of SAS Closed compressor C-9008 115928 Dewpoint rises when chambers IAS Open switch 116248 Investigate discharge valve SAS Closed on C-900C 116776 Evaluate cause of air SAS Open compressor trip 116254 Problems with regeneration IAS Open of beds - Y-910C 117078 Transfer valve sticks on IAS Closed Y-910A and B Resolution: Implement the subject work requests as soon as scheduling permits. Closure Requirement: The hrs listed above. Completion Date: Initial plant heat up. 4C.32.9 Problem 9 IE Informational Notice No. 86-51 4C.32-4
. Tracking Number: 26.0116 f%
Description:
IE Notice No. 86-51 is concerned with potential problems in the pneumatic supply lines, specifically those near the accumulators serving automatic depressurization systems (ADSs) in BWRs due to excessive leakage. Resolution: Nuclear Engineering to review the subject IE Notices and Bulletins regarding new and existing valves. Closure Requirement: QCI-12 closure form Completion Date: Plant heat up. 4C.32.10 Problem 10 The Turbine Bypass Valves (TBVs) fail closed upon a loss of IAS. The accumulator tanks are not to coc'e. Tracking Number: 26.0343
Description:
Upon loss of instrument air, the TBVs use their associated accumulators to supply air to hold the valves closed. While on accumulator air, both Control Room and ICS valve control is inhibited. The ' accumulators do not meet the applicable code.
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Resolution: Install component level air bottle back-up. Remove existing accumulators. Closure Requirement: ECN R0859 Completion Date: Prior to heat up 4C.32.11 Problem 11 Operator training is needed due to system modifications and the revisions to the operating procedures. , Tracking Number: 26.0118
Description:
Operator training is deficient due to new system modifications and the revisions to the operations procedures. ( Resolution: Provide training for operators on new modifications and l procedures. Closure Requirement: Operators complete requalification training on procedures and modifications. 4C.32-5 l
Completion Date: Prior to heat up. 4C.32.12 Problem 12 Upon loss of control power to the plant air compressors, they unload and TCW is isolated. Tracking Number: 26.0348
== Description:== Upon a loss of Bus Voltage, the at
- mpressors unload (de-energized state of solenoid) a me TCH is isolated from the compressor cooli , water jackets. In addition, the compressor vendor re ammends a " slight constant trickle of water across the thermostatic bulb:
after the compressor is shut down, which the present valve lineup does not permit. Resolution: Revise Operating Procedure A.40, and casualty procedure for the loss of IE bus. Closure Requirement: Approval of procedure revisions Completion Date: Initial plant heat up. O O 4C.32-6
4C.33 Auxiliary Feedwater System (AFWS) O 4C.33.1 Problem 1 The requirements of IE Bulletin 85-03 has not been satisfied. Tracking Number: 19.0038
Description:
The NRC issued IE Bulletin 85-03, in late 1985, mandating development and implementation of a program to ensure that Motor Operated Valves (MOVs) belonging to the (HPI) High pressure Injection and AFW systems which are required for operational readiness have their valve operator switches, selected, set tested, and properly maintained. Resolution: Implement the requirements of IE Bulletin 85-03 as identified on ECN R-0914. The following actions will be completed:
- 1) Collect and establish current valve conditions.
- 2) Refurbish operators.
- 3) Analytically determine torque and thrust.
- 4) Confirm wiring material and arrangement.
- 5) Set switches using analytical values.
- 6) Test to confirm the valve will perform when required (MOVATS).
l 7) Revise procedures, if required. Closure Requirement: ECN R-0914 and associated testing. Completion Date: Prior to Unit S/0 Turnover to Dispatcher. 4C.33.2 Problem 2 Power supply not uniform to all equipment Tracking Number: 26.0119
Description:
AFW pump suction and discharge pressure transmitters not powered by same power source as pumps they monitor. l Resolution: Modify power supply to the instruments per ECN A-54150. l 4C.33-1 l 1
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Testing: Perform instrument calibration procedure I.038 on PT-31801, 31803, 31901, and 31903. Closure Requirement: ECN A-5415Q and instrument calibration I.038 on PT-31801, PT-31803, PT-31901 and PT-31903. Completion Date: Prior to initial plant heat up. 4C.33.3 Problem 3 Component Level Instrument Air Backup Tracking Number: 18.0037
Description:
Component level air backup are needed for the AFH Control Valves in order to provide instrument air in the event of a loss of normal air supply. Resolution: Install component level backup air bottles (ECN R-0859) for valves FV-20527 and FV-20528. Closurc Requirement: ECN R-0859 Completion Date: Prior to Loss of Offsite Power Test and ESF Actuation Testing. 4C.33.4 Problem 4 District must comply with NUREG 0737 relating to automatic AFW initiation and indication. 1 Tracking Number: 18.0031
Description:
As a result of the NRC's TMI action plan, the District was required to comply with AFH upgrades. Investigation: In NUREG 1195 the NRC concluded that post-THI requirement II.E.1.1 specifies that one train of AFH be operable for postulated loss of main feedwater complicated by loss of offsite and onsite AC power sources. In such a case, the AFH could have to function on other power sources, such as steam and DC (i.e., battery) power. In addition, the NRC concluded that the turbine-driven pump will operate only if the steam supply / isolation valve is actually in the open position. Also, the AFH (SFAS) isolation valve will remain closed without AC electric power. AFH flow then depends upon the AFH (ICS) flow control valve. O 4C.33-2 l
Resolution: The installation of EFIC ECN A-5415 will modify and p upgrade controls and power supplies for valves. ; l Closure Requirement: ECN A-5415 '
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l Completion Date: Prior to heat up. 4C.33.5 Problem 5 Present piping configuration does not allow test 9 site reservoir system suction supply to AFW supply. Tracking Number: 26.0126
Description:
The AFH system is designed to take a suction on the CST during normal operations. If the Condensate Storage Tank (CST) would no longer be capable of supplying demineralized water to the suction of the AFW pump, the piping configuration allows for an alternate supply of water from the SRS system. Either the site reservoir or the Folsom south canal would be able to supply the required makeup. Testing the alternate supply of water, untreated " RAH" water, is not given consideration due to the present piping configuration. Therefore, there is no method to provide Site Reservoir system (SRS) to AFW flow verification during testing. V Resolution: Modify piping configuration temporarily to allow testing the capability of the alternate water supply. Closure Requirement: Perform a STP to demonstrate that the SRS system provides adequate flow and NPSH for the AFH pumps. Completion Date: Prior to initial plant heat up. 4C.33.6 Problem 6 The AFW System may not comply with licensing commitments. Tracking Number: 17.00018
Description:
The AFW system should be upgraded to withstand an external event such as wind and turbine generated missiles, a design basis event, and Appendix R fire, a high energy line break, and meet Quality Class II over I Seismic requirements.
- Resolution
- Modify the AFW system LAW Licensing commitments. -
O 4C.33-3
Closure Requirement: ECN A-5415 Completion Date: Prior to reactor start up. 4C.33.7 Problem 7 Manual Operation of AFW Control Valves with Flow Tracking Number: 15.0173
== Description:== The failure of FV-20527 during the December 26, 1985 transient resulted in the uncertainty of the adequacy of the local manual operator under various conditions. The handwheel assembly is designed to be operated without a cheater. During the transient a cheater was used on FV-20527 which cause handwheel assembly damage. Resolution: Determine the local control operability of FV-20527 and FV-20528 with both AFW pumps running and full Delta P, one AFW pump running, operate valve with local manual operator. Closure Requirement: Develop and perform a STP to verify that FV-20527 and FV-20528 can be operated manually during high differential pressure conditions during AFH system flow testing. Completion Date: Prior to initial plant heat up. 4C.33.8 Problem 8 Handwheel Assembly Position Indicator Tracking Number: 15.0042
== Description:== The valve position labels on auxiliary feedwater flow control valves (FV-20527 and FV-20528) are not aligned with the valve position pointers. Resolution: ECN R-0470 has been implemented to remove the existing label for the hand operator and replaced it with a label that indicates the quantitative position of valve hand operator. Closure Requirement: Perform an inspection of the valve handwheel indicator. Completion Date: Prior to initial plant heat up. O 4C.33-4
4C.33.9 Problem 9 AFW Pump Turbine may overspeed during start up Tracking Number: 26.0061
Description:
During a surveillance procedure test results review, it was revealed that steam inlet valve (SFV-30801) has a stroke opening time of approximately 8 seconds after modifications and overhaul. Previously this valve operated in approximately 18 seconds. Resolution: Testing of the Terry Turbine. Closure Requirement: Test-results must be satisfactory. Completion Date: Prior to Unit S/U release to Dispatcher. 4C.33.10 Problem 10 AFW Pump Turbine Governor Valve Seat Cracks Tracking Number: 26.0062
Description:
During maintenance inspections at Turkey Point Units 3 and 4, cracks were discovered in the AFH pump turbine O governor valve upper seats. Investigate the , construction and materials used for Rancho Seco AFH turbine governor valve. Resolution: (Disassemble, inspect, and repair the Govenor Valve, if required.) Closure Requirement: Inspection results satisfactory. Completion Date: Prior to initial plant heat up. l 4C.33.11
- Problem 11 AFW Flow Control Valve Sten Position Indication is not visable during manual operation.
Tracking Number: 26.0223 i i 4C.33-5
== Description:== The valve stem position on the AFH flow control valves FV-20527 and FV-20528 cannot be easily utilized during manual valve operations. Resolution: A new valve stem position Indication system to be designed and installed on FV-20527 and FV-20528. Closure Requirement: Complete work request. Completion Date: Prior to heat up. 40.33.12 Problem 12 Turbine driven AFW pump flow may not meet T.S. 5 requirements. Tracking Number: 26.0065
== Description:== Previous surveillance procedures have not required the turbine driven AFH pump flow to meet the current = technical specifications requirements; therefore, the applicable surveillance procedure must be revised. Resolution: Revise SP.210.01A, monthly turbine / motor driven auxiliary feed pump P-318 surveillance and inservice test. Closure Requirement: Perform SP 210.01A per technical specifications. (Section 4, function a). Completion Date: Prior to initial plant heat up. 4C.33.13 Problem 13 AFW System Operator Training must be accomplished before restart. Tracking Number: 26.0609
== Description:== Major modifications have been made to the AFH control system. Particular emphasis should be placed on the significance of initiating steam and feed flow isolation, as well as the new controllers installed by EFIC. O 4C.33-6
Resolution: Conduct training. The operator training program to be
>Q changed to include the EFIC modification and an AFW ,Q Mods.
Closure Requirement: Operator training records. Completion Date: Prior to reactor start up. 4C.33.14 Problem 14 AFW Valves must be tested at worst case conditions Tracking Number: 26.0066
Description:
As a result of PRC review of the detailed account of the Davis-Besse-1 transient, the PRC proposed an investigation be implemented to determine if AFH valves are tested at " worst case" condition; i.e., during maximum flow and differential pressure conditions.
! Resolution: Test valves at worst case conditions.
Closure Requirement: Develop and perform a special valve stroke test to demonstrate the following valves will operate during maximum differential pressure and full flow conditions. HV-31827 HV-31826 HV-20577 HV-20578 HV-20581
, HV-20582 SFV-30801 Completion Date: Prior to Unit S/U Turnover to Dispatch.
4C.33.15 Problem 15 Discrepancies exist between Vendor Tech Manual and SMUD Procedures for Maximum Allowable Number of AFW Pump Starts l Tracking Number: 26.0067 i 4C.33-7
== Description:== Discrepancies were noted between AFW pump motor starts allowed per vendor technical manual and the limits and precautions in the operating and surveillance procedures. Resolution: Determine correct pump start limit and update the incorrect reference (s). Closure Requirement: A demonstration that AFW pump motor windings do r,ot exceed design temperature after multiple starts will be performed. An AST will be executed. Completion Date: Prior to initial plant heat up. 4C.33.16 Problem 16 AFW Mods must be incorporated into Operating Procedures Tracking Number: 15.0147.A
== Description:== Due to modifications of AFW valve controls the affected Operating, Casualty and Emergency procedures must be revised. Resolution: Adoption of new procedures. Closure Requirement: New procedures. Completion Date: Prior to reactor start up. 40.33.17 Problem 17 AFW pump operability may be adversely affected by Deluge System Tracking Number: 19.0035
== Description:== An AFW pump area deluge test has been proposed per STP-1004. It is recommended that this test be completed prior to start up to comply with a commitment to the NRC and to clearly demonstrate that operation of the deluge system will not cause an AFW system failure. Resolution: Perform engineering study of possible design changes for motor protection during deluge spray. Functional test as required. O 4C.33-8
7 ., - Closure Requirement: Complete study, modifications and testing as required. " Completion Date: Prior to reactor start up. r, - 4C.33.18 Problem 18 Unreliable communications between the Control Room and AFW control valves area may exist. S- Tracking Number: 26.0087
Description:
During previous conditions which have required use of the AFW Control Valves' sound powered phones, the communications between the valve area and the Control Room have been unreliable. Resolution: Upgrade communications equipment. Closure Requirement: A check of communications will be performed during flow i testing of the AFW system. Completion Date: Prior to initial plant heat up. 4 4C.33.19
; Problem 19 Revise operating procedure to mandate stop of AFW pumps y If fIow cannot be secured in excess feed situations.
Tracking Number: 15.0146
Description:
Plant Operations Emergency Procedure E.05, " Excessive heat Transfer," Step 2 and 3 should be rewritten to advise the operator to stop the associated AFH pump.
- ( If the AFH flow cannot be stopped using the controls in the Control Room the auxiliary feedwater pump should be stopped to prevent overcooling.
Resolution: Emergency Procedure E.05 to be revised. Closure Requirement: E.05 revision. Completion Date: Prior to initial plant heat up. I I b= ~ ) O i , 4C.33-9 m- _. __ - - -
4C.33.20 Problem 20 Turbine trip testing has not been performed. Tracking Number: 26.0608
== Description:== Terry turbine overspeed trip testing has not been performed. Also, a terry turbine " start up after a short shutdown test" would demon:;trate that the terry turbine can be restarted after is has been shutdown for a minute or so. Resolution: Perform tarry turbine overspeed trip testing prior to start up and that overspeed testing is added to the routine PM Program. Closure Requirement: Perform a STP to demonstrate the AFH turbine overspeed trip setpoint and start up after trip capability. Completion Date: Prior to reactor start up. 4C.33.21 Problem 21 Open AFW Work Requests Trarking Number: 26.0091 Detcription: A review of open work requests was performed to determine the work requests required to be closed out prior to start up. Resolution: Complete all open work requests on the AFW system. Closure Requirement: Complete all work requests as necessary and testing specified. Completion Date: Prior to reactor start up. 4C.33.22 Problem 22 Condensate Storage Tank Instrument Upgrade Tracking Number: 26.0121
== Description:== No Class 1 instrumentation exists in the Control Room for the CST. Resolution: Install Class 1 CST level instrumentation per ECN R-0952 which inputs to SPDS. Closure Requirement: ECN R-0952 completed. 4C.33-10
Completion Date: Prior to initial plant heat up. 40.33.23 Problem 23 AFW pump discharge pressure instruments should be upgraded. Tracking Number: 26.0204
Description:
Instrumentation does not exist in the Control Room for AFH pump discharge pressure indications. Resolution: Installation of EFIC ECN A-5415. Closure Requirement: ECN A-5415 Completion Date: Prior to heat up. 4C.33.24 Problem 24 AFW pump bearing temperature instrument should be upgraded. Tracklag Number: 26.0122 O
Description:
Instrumentation does not exist in the Control room in order to monitor AFH pump bearing temperature. Resolution: Install instrumentation to provide the Control Room AFW pump bearing temperature indication. Closure Requirement: Complete ECN and revise SP210.01A and SP210.018. Completion Date: Prior to initial plant heat up. 4C.33.25 Problem 25 SFAS Valves go fully open upon SFAS initiation. Tracking Number: 26.0185 l
Description:
Upon SFAS actuation, valves SFV-20577(78) go fully open resulting in a potential for RCS overcooling. Resolution: Installation of EFIC, ECN A-5415. Closure Requirement: ECN A-5415 O 4C.33-11
Completion Date: Prior to initial plant heat up. 4C.33.26 Problem 26 Need to instali controls for AFW Valves at shutdown panel H2SD Tracking Number: 15.0181
Description:
Controls for AFW valves FV-20527(28) do not exist outside the Control Room. Alternate shutdown capability for the AFH system outside the Control Room is required in order to be in compliance with 10CFR50, Appendix R. Resolution: ECN A-5415 will install controls at H2SD Panel. Closure Requirement: ECN A-5415. Completion Date: Prior to initial plant heat up. 4C.33.27 Problem 27 AFW valve controls are located on control panels which are not near one another. Tracking Number: 26.0186 Description; "A" OTSG valve (SFV-20577) controls are located on"B" SFAS panel and "B" SFAS panel and "B" OTSG valve (SFV-20578) controls are located on "A" SFAS panel. Installation of EFIC ECN A-5415. Resolution: Closure Requirement: ECN A-5415 Completion Date: Prior to initial plant heat up. l 40.33.28 Problem 28 AFW Valve controls are not arranged for convenient and logical operation action. Tracking Number: 21.0004.A
Description:
Leeds & Northrup controllers for AFW valves FV-20527(28) can only be operated in manual mode because they are designed to be independent of ICS. O 4C.33-12
Resolution: The installation of EFIC ECN A-5415 will replace these G controllers and provide for manual as well as automatic control of AFW flow. Closure Requirement: ECN A-5415 Completion Date: Prior to initial plant heat up. 4C.33.29 Problem 29 The AFW system is controlled by non-safety systems. Tracking Number: 26.0187
Description:
Integrated Control System (ICS) and Non-nuclear. Instrumentation (NNI) have input to control (s) of AFW system. Resolution: EFIC ECN A-5415 will install a Class I control system for AFW valves and pumps. Closure Requirement: ECN A-5415. Completion Date: Prior to initial plant hsat up. 4C.33.30 Problem 30 Auxiliary feedpumps may have been damaged during the December 26, 1985 transient. Tracking Number: 15.0282
Description:
During the December 26, 1985 transient the auxiliary feedpumps were operated in run out condition for several minutes. The condition of the pump's internals l and performance characteristics of the pumps are j indeterminate. Resolution: Perform vendor evaluation or test pump per vendor specs. Closure Requirement: Perform SP210.01A and SP210.018 and obtain the appropriate vendor documentation. O 4C.33-13
Npletion Date: Prior to initial plant heat up. 4C.33.31 Problem 31 Loss of Instrument Air (IAS) may cause overcooling accident. Tracking Number: 20.0166
Description:
QA 82-19 states the most serious consequence of a total loss of the Instrument Air System (IAS) would result in a possible severe overcooling transient from uncontrolled AFW flow. Resolution: Installation of EFIC ECN A-5415 will make IAS more reliable. Closure Requirement: ECN A-5415 Completion Date: Prior to initial plant heat up. 4C.33.32 Problem 32 Water slugs may inadvertently enter the AFW Turbine. Tracking Number: 26.0287
Description:
Terry turbine recommends for K-308 that a steam separator of the proper size, with a redundant trap of ample capacity be installed in the turbine inlet piping to minimize the potential and effects of water slugs on the turbine. No evidence of steam separator exists in l the current inlet piping configuration or of an l analysis by Nuclear Engineering to justify the ability l of the current inlet piping configuration to minimize l the potential and effects of water slugs. Resolution: Provide steam trap modifications as necessary to I minimize the probability of water slugs entering K-308. ECN A-3062. Closure Requirement: ECN A-3062. Completion Date: Prior to initial plant heat up. i 4C.33.33 Problem 33 Thermal binding of AFW gate valves may occur. Tracking Number: 26.0951 4C.33-14
Description:
SOER 84-7 addressed the susceptibility of safety p related gate valves to thermal binding pressure locking and bonnet overpressurization. Resolution: Modify SFV-20577 (CCL #T860428452C) and other valves as required. I Closure Requirement: Specify and implement the required hardware modification. Completion Date: Prior to reactor start up. 4C.33.34 Problem 34 Install AFW flow control station in Control Room. Tracking Number: 26.0953 l
Description:
- Provide AFH flow control station in control room to l position control valves upon loss of ICS power and independent of ICS, also provide OTSG level indication (narrow range).
Resolution: EFIC ECN A-5415 will provide installation of the proper components. Closure Requirement: ECN A-5415 Completion Date: Prior to Unit S/0 Turnover to Dispatcher. 4C.33.35 Problem 35 Casualty Procedures are not clear concerning operator guidance on when to trip the Main Feedwater Pumps. Tracking Number: 22.0408
Description:
Casualty Procedure C.10. " Loss of Steam Generator Feed" should be revised. Resolution: Revise Casualty Procedure C.10 to agree with emergency Procedure E.05, " Excessive Heat Transfer." Closure Requirement: Revision of Casualty Procedure C.10. O 4C.33-15
Completion Date: Prior to initial plant heat up. 4C.33.36 Problem 36 AFW pump flow calculations should be reviewed to ensure they provide required head. Tracking Number: 26.0952
Description:
Installation of EFIC will require calculations to determine flows to steam generators. Resolution: Perform calculations. The appropriate assumptions should include the suction source, accident analysis, and flow paths utilized. Closure Requirement: Complete AFH pump performance calculations, revised performance calculations, revised procedures SP210.01A and SP210.018. Completion Date: Prior to initial plant heat up. l ( l O l l O 4C.33-16 i
-- p
4C.34 NSEB Normal and Essential HVAC System (NHVS) O 4C .34.1 i Problem 1 Temporary Air Flow Restrictions in the form of sheet metal blanks have been installed in the NSEB Essential Air Handler filter banks. Tracking Number: 19.0010I
Description:
Drive sheaves in the NSEB Essential Air Handlers have been oversized. These oversized sheaves cause the fan to move an excessive amount of air. Sheet metal l blank-off sheets have been temporarily installed to i increase system airflow resistance, thereby reducing the flow rate. Resolution: Revise NSEB Essential Air Handler air flow rate to that specifled in technical Specifications. Remove sheet metal blank-off plates. Revise SP .0085A and SP .00858 Closure Requirement: Technical Specification 4.31; NCR 4790, SP .0085A, and s SP .00858, ECN R 1119. l'\ Completion Date: Prior to initial plant heat up. 4C.34.2 ' Problem 2 The thermostat which controls operation of the NSEB Essential Air System may have to be moved. Tracking Number: 19.0010F
Description:
An inverter and a battery charger are located in NSEB Room 232. Heat generated by this equipment causes this to be the warmest room in the NSEB. The thermostat which controls cooling is located in a separate room. Due to this, temperature in Room 232 is subject to exceeding the setpoint temperature. There is a possibility that the electrical equipment in Room 232 could be damaged by the elevated temperature. After essential system actuation, the air handler, which is controlled by the thermostat, does not operate continuously. 4C.34-1
Resolution: Revise the setpoint of the existing thermostat or install a thermostat in Room 232. Closure Requirement: Process Standard AP 154 ECN R 1107 Completion Date: Prior to post modification testing (approximately February 15, 1987) 40.34.3 Problem 3 Cooling capability of the NSEB Essential Air System is marginal during cold weather operation. Tracking Number: 26.0418
== Description:== The controls of the refrigeration portion of the HVAC system do not permit reasonable flexibility of the system to respond to all variations in cooling loads with respect to ambient conditions. In particular, during low ambient conditions, the system falls to maintain adequate pressure in the suction line, resulting in compressor trips. The consequence of this operation is that the essential HVAC system cannot be relied upon to operate under all postulated accident modes. Resolution: Revise the controls for the condenser fans to allow cycling of two of the three fans. Closure Requirement: ECN R 1026, ECN R 1023, STP 1066, STP 1067. Completion Date: Prior to initial plant heat up. 40.34.4 Problem 4 Unloader solenoid valves on compressors in Essential NSEB HVAC units U-503 A/B vibrate excessively with unit in operation. Tracking Number: 26.0419
== Description:== Solenoid manifold wiring is installed in galvanized pipe without proper flex connections and is bolted to metal deck plate of compressor cabinets. Resolution: Replace part of hard pipe with flexible tubing to reduce vibration coming from deck plate. Closure Requirement: Need nuclear engineer evaluation. 4C.34-2
Completion Date: Prior to post modification testing (approximately February 15, 1987) 4C.34.5 Problem 5 Fire Dampers 39N and SON were installed without ETLs in the NSEB. Tracking Number: 26.0421
Description:
ECN A-3725A was issued to install ETLs on FDs 39N and SON of the NSEB. Resolution: Install per ECN A-3725A Closure Requirement: ECN A-3725A Completion Date: Prior to initial plant heat up. 4C.34.6 Problem 6 NSEB Essential HVAC System Refrigeration Problems. Tracking Number: 26.0422 O
Description:
Modifications are necessary to facilitate efficient and effective Refrigeration System operation. Resolution: Install modification to provide refrigeration system pumpdown capability, hot gas bypass control, evaporator control, and compressor control per ECN 1023. Closure Requirement: ECN 1023, STP 1066, STP 1067. Completion Date: Prior to initial plant heat up. l 40.34.7 Problem 7 The Essential refrigeration component arrangement does < not allow convenient maintenance. Tracking Number: 26.0423
Description:
The arrangement of the refrigeration system components does not facilitate malatenance of the system due to:
- a. insufficient isolation capability,
- b. difficult access to components, and
- c. extreme difficulty in removing major components for repair or replacement.
4C.34-3
Resolution: Write a memo evaluation proposed modification. Install liquid line sight glass per ECN-R-1158 Install receiver sight glass per ECN-R-1160. Closure Requirement: ECN complete. . Completion Date: Approximately February 15, 1987. 4C.34.8 Problem 8 The NSEB Essential Refrigerant Compressor B has suffered severe damage and must be rebuilt or replaced. Tracking Number: 26.0426
Description:
The refrigerant compressor exhibits scored cylinder walls, failed valves, and damaged pistons. Replacement of cylinder liners, pistons, valves, valve springs, and miscellaneous items will be necessary to restore the compressor to operable condition. Resolution: Install a replacement compressor. Initiate compressor failure report. Provide a spare compressor. Closure Requirement: Install rebuilt or spare compressor. Compressor failure report. Spare compressor purchased. Completion Date: Prior to initial plant heat up. 4C.34.9 Problem 9 Various rooms in the NSEB are improperly pressurized i when either of the NSEB Normal or Essential Air l Conditioning Systems is in operation. t l Tracking Number: 19.0010H
Description:
The NSEB Normal and Essential Air Conditioning air i distribution systems have not been properly balanced. l Certain rooms in the NSEB have been balanced so that supply air flow exceeds return air flow. This condition results in room pressurization to achieve equilibrium. Due to this condition, access doors are difficult to open or close; door locking mechanisms do l not operate properly and have incurred damage; and fire door closing devices do not function properly. O l 4C.34-4
}
Resolution: Balance the NSEB Normal and Essential Air Conditioning Systems. V Closure Requirement: Work Request Complete. Completion Date: Prior to initial plant heat up. 4C.34.10 Problem 10 Operation of the NSEB Essential Air Refrigerant Systems is compromised by certain. piping and control problems. Tracking Number: 26.0427
Description:
A leak in the NSEB Essential Air Train A refrigerant piping is suspected. Evaporator coil superheat should be checked and adjusted. Resolution: Repair identified refrigerant piping leaks, WRN 117987. Adjust evaporator superheat. Closure Requirement: Closed Work Requests, HRN 117987. Completion Date: Prior to post modification testing (approximately February 15, 1987). 4C.34.11 Problem 11 Plant Operating Procedure A.14 does not provide adequate information on the location of electrical breakers and proper position / mode correlation for the NSEB air conditioning system isolation dampers. Tracking Number: 26.0431
Description:
Plant Operating Procedure A.14 does not include a listing of the power supply breakers for the following dampers: HV-50104 - NSEB B Train Normal Train Supply Air Damper HV-50105 - NSEB A Train Normal Train Supply Air Damper HV-50126 - NSEB B Train Normal Train Supply Air Damper HV-50127 - NSEB A Train Normal Train Supply Air Damper HV-50135 - NSEB A Train Essential Unit Return Air Damper HV-50136 - NSEB B Train Essential Unit Return Air Damper l i HV-50137 - NSEB A Train Essential Unit Outside Air Isolation Damper HV-50158 - NSEB B Train Essential Unit Outside Air Isolation Damper. 4C.34-5
1 HV-50188 - NSEB B Train Essential Unit Supply Air Isolation Damper . HV-50189 - NSEB A Train Essential Unit Supply Air l Isolation Damper i HV-55307 - NSEB A Train Normal Unit Return Air l Isolation Damper ! HV-55310 - NSEB B Train Normal Unit Return Air , Isolation Damper l Resolution: Revise Operating Procedure A.14 Closure Requirement: Operating Procedure A.14 Completion Date: Prior to initial plant heat up. l 4C.34.12 Problem 12 NSEB Essential HVAC Refrigeration System P&lD. Tracking Number: 26.0432
Description:
A P&ID drawing depicting the NSEB Essential Air Refrigeration System does not exist. A drawing is necessary to facilitate operation, maintenance, and training. Resolution: Prepare a new P&lD drawing illustrating the NSEB essential air refrigeration system. Closure Requirement: P & ID approved. Completion Date: Prior to initial plant heat up. l 4C.34.13 i Problem 13 Procedures and training, which would facilitate maintenance on an expedited basis for the NSEB l Essential Air Refrigerant System do not exist. Tracking Number: 26.0433
Description:
Procedures and training which would facilitate maintenance on an expedited basis do not exist. Consequently, replacement and/or repair of a major component canrot be accomplished in an efficient manner. This situation could cause an unscheduled station outage or extend the length of a scheduled outage. l Resolution: Write maintenance procedure. l Provide P & 10. Provide training for maintenance personnel. l 4C.34-6 l -
Closure Requirement: Procedure, P & ID, and training provided. Complet!nn Date: Prior to initial plant heat up. 4C.34.14 Problem 14 Temperature in the NSEB Essential Unit compressor l compartments may become excessive during hot weather operation. ~ Tracking Number: 26.0437
Description:
Ventilation in Essential Unit compressor compartments is provided through natural circulation. During hot weather operation, temperature in the compartments is elevated when the system is in operation. The elevated temperatures may result in failure of the motor, compressor, or other component. Resolution: Modify the Essential Unit compressor compartments by > providing a vent between the compressor compartment and the condenser plenum, per ECN-R-1170. Closure Requirement: ECN completed. Completion Date: Prior to post modification testing. 4C.34.15 Problem 15 Temperature controllers for the normal HVAC systems should be relocated. Tracking Number: 26.0438
Description:
The temperature controllers for the NSEB normal HVAC are located in the return ducting to the unit. Mixing of return air does not allow temperature control for the inverter rooms, which are the hottest areas in the NSEB. Resolution: install thermostats per ECN R1107 Closure Requirement: ECN R 1107 Completion Date: Prior to post modification testing. a 4C.34-7 l
4C.34.16 Problem 16 Fire dampers for the essential HVAC System not closing when actuated. Tracking Number: 26.0439
== Description:== Fire dampers will not close completely due to binding in the channels and possible other causes. Resolution: Inspect fire dampers Review test requirements Closure Requirement: Satisfactory test results of fire dampers. Completion Date: Prior to initial plant heat up. O O 4C.34-8
4C.35 Reactor Coolant Drain System (RCDS) V 4C.35.1 Problem 1 Review of open RCD work requests. Tracking Number: 26.0601
Description:
Open work requests were reviewed for impact on Restart. One item has been worked, awaits paper closecut. Other items relate to calibrating a level transmitter and adjusting a limit switch. Resolution: Calibrate LIT-20510 reference level to full scale of 120 inches. Adjust SFV-60004-W striker arm to make contact. Closure Requirement: Work request numbers 117589, 123146, 110064 Completion Date: Prior to leaving cold shutdown. 4C.35.2 Problem 2 Review of open Work Requests for the Reactor Coolant Drain System. V Tracking Number: 26.0598
Description:
Plant Maintenance report review for open work documents not previously identified, for impact on restart. Specific items relate to calibration of various LE's and LIT's. Resolution: Complete work requests. Closure Requirement: Work request numbers 119870, 871, 873, 874. Completion Date: Prior to leaving cold shutdown. O O 4C.35-1
4C.36 Drainage and Sewerage System (CDS) 4C.36.1 Problem 1 Review of open Work Requests, CDS Tracking Number: 26.0695
Description:
Reviewed plant maintenance report of open work requests for impact on restart. Items relate to inoperable sample pump, leaking recirc. valve and removal of a temporary drain line. Resolution: Repair sample pump P-687. Repair recirc. valve CDS-511. Remove DRCST transfer Iine. Closure Requirement: 14ork request numbers 106893, 114910, 123347 Completion Date: Prior to leaving cold shutdown. 4C.36.2 Problem 2 Possibility of site contamination and Tech Spec violation, CDS. Tracking Number: 15.0024
Description:
Make necessary changes to procedure to immediately divert the plant effluent to the retention basin and L secure cooling tower blowdown if water is in the main steam lines and a situation exists wherein the code safeties or ADV (Atmospheric Dump Valve) may lift. I Resolution: Write new E.06. Closure Requirement: Adopt new procedure. Completion Date: Prior to leaving cold shutdown. l l l 4C.36-1
4C.37 Radwaste System (RWS) 4C.37.1 Problem 1 Capacity of secondary waste water system. Tracking Number: 19.0040.A
Description:
Storage and processing of waste water (secondary side) is currently too limited. Low levels of contamination on the secondary side sometimes exist. This modification provides for a new regenerant hold-up tank (RHUT), new makeup demin sump, and other pathways to separate clean waters from radwaste. Resolution: Install new equipment and make operational items associated with ECN R-0775. Closure Requirement: ECN-R-0775 Completion Date: Prior to leaving cold shutdown. 4C.37.2 Problem 2 The classification of liquid effluent as " radioactive liquid effluent" needs to be clearly defined. Tracking Number: 21.0283
Description:
The classification of liquid effluent as " radioactive 11guld effluent" needs to be clearly defined to meet all regulatory and license requirements, balanced against performing radiochemical analysis for extremely low levels of radioactivity that go well beyond the ALARA concept and are at or below environmental levels of radioactivity. Resolution: 1) DGM, Nuclear conduct a press conference to clarify to the public that Rancho Seco does discharge very small quantities of radioactive material in liquid effluents that are well below all regulatory requirements.
- 2) The Radioactive Protection Group and Radiochemistry Group revise the radioactive liquid effluent control procedures to improve the radiochemical analysis, accountability, and administrative controls.
- 3) An amendment to the Rancho Seco Technical Specifications be prepared and submitted to the NRC C that clarifies the Limiting Conditions for 4C.37-1
l l Operation and the Surveillance Requirements related to radioactive !Iquid offluents and how Rancho Seco wiII be operated within the numerical guides for design objectives in 10CFRS0, Appendix 1. Closure Requirement: 1) Press Conference. ;
- 2) Procedure Changes.
- 3) Proposal Amendment to the Technical Specifications.
Completion Date: 1) September 8, 1986.
- 2) October 8, 1986.
- 3) Submitted to the NRC prior to start up.
4C.37.3 Problem 3 Evaluation shall be done of testing liquid release pathways. Tracking Number: 26.0307
== Description:== To ensure the integrity of all 11guld effluent release pathways, existing surveillance procedures and other testing requirements will be reviewed to ensure that all testing is covered prior to start up. Most of the testing will be performed while performing start up functions under M00530, ECN R-0775, Problem No. 2. Resolution: Investigative Reports. Closure Requirement- Memo from Systems Engineer with results of review. Completion Date: Prior to leaving cold shutdown. 4C.37.4 Problem 4 Evaluation of modifications needed for integration of primary and secondary waste water processing. Tracking Number: 19.0040.B Priority: 1
== Description:==
- 1) Add sump in spent fuel pool cooler area to collect rad water leakage or pipe break.
O 4C.37-2 , I
- 2) Make main steam line drains pump to this sump.
- 3) Route overflow from T-993 0 (misc. water hold-up tank) to this sump and alarm high level in control room.
- 4) Replace P-376 with larger pump.
- 5) Route T-375 (condensate return tank) overflow to-radwaste system to prevent spills to the Aux. Bldg.
floor during periods of both evaporators operating. Ilesolution: Nuclear Engineering to study ar.d evaltar.. the above and other potential unmonitored releases as criority 1. Closure Requirement: Results of Nuclear Engineering Study and ECN's required. Completion Date: Start up. O O 4C.37-3
4C.38 Control Rod Drive (CRD) O 4C.38.1 Problem 1 CRD procedure should be modified. Tracking Number: 26.0755
Description:
Three modifications required are:
- 1) Procedure B.4 to check AP.101.14 (temperature -
pressure limits) prior to withdrawing rods.
- 2) Emergency Operating Procedure (EOP) Rule 6 (2.1.3) to verify A.74, RCS temperature / pressure requirements are satisfied prior to control rod drive operation.
- 3) Operating Procedure A.1 is to have a precaution against tripping the reactor with rods withdrawn if a condition requiring RCS venting occurs. These conditions are listed in OP A.1 Step 3.8 and Technical Specification 3.1.9.
Resolution: New Procedures. Closure Requirement: Adopting new procedures. Completion Date: Restart. l O 4C.38-1
4C.39 Waste Gas System (WGS) 4C.39.1 Problem 1 The make-up (surge) tank could not be sampled at power due to leakage in the system. Tracking Number: 21.0102D
Description:
The auxiliary building stack monitor would alarm when the sample line was lined up. This is an ALARA and Technical Specifications release concern. Resolution: Perform SP-211.02B to locate leaks in the waste gas system. Closure Requirement: Completed. Completion Date: Complete. 4C.39.2 Problem 2 Review of open work request. Tracking Number: 26.0597
Description:
Review of open work request to determine the impact on restart. Found: HR 106787, V-650, waste gas surge tank, plugged line or drain tap. HR 119047, C-651A, a waste gas compressor, investigate and repair inoperable compressor. Resolution: Maintenance to include these items into their restart schedule. Closure Requirement: Completed. Completion Date: Complete. O , 4C.39-1
4C.40 Balance of Plant Systems (OHVS) 4C.40.1 Problem 1 Casualty Procedures for Heating, Ventilating and Air Conditioning do not meet T.S. Requirements Tracking Number: 22.0256
Description:
Table 3.5.1 requires Reactor Building purge valves SFV-53504, SFV-53604, and SFV-53610 to be closed and their respective breakers opened during events where they are declared inoperable. Resolution: Revise Procedure C.28 to include the requirements of TS Table 3.5.1-1. Closure Requirement: None Completion Date: Prior to reactor heat up. 4C.40.2 l Problem 2 Mounting supports for exhaust fans EFA 1 through 4 do not meet seismic requirements. Tracking Number: 26.0879 l J
Description:
Exhaust fans (EFA-1 through 4) mounting support required upgrading to meet seismic class 1 requirements. Resolution: Impiement ECN: R-1002 Closure Requirement: ECN R-1002 Completion Date: Prior to reactor heat up. l I s 4C.40-1
1 4C.41 Reactor Sampiing System (RSS) O 4C.41.1 Problem 1 RSS heat tracing does not function properly. Tracking Number: 26.0461
Description:
The heat tracing on the RSS does not appear to function properly. The heat tracing temperature controllers do not maintain temperatures within the proper range. Resolution: Modify the heat tracing system and lagging to correct the problem. Closure Requirement: ECN R-0488; NCRs S-5267 and S-5268; and WRs 112042, 112043, 112061, and 112062. Completion Date: Prior to PASS acceptance test. 4C.41.2 Problem 2 The PASS system does not work properly. Tracking Number: 26.0462
Description:
The PASS does not work properly, is poorly designed, and is located in such a way that it may not be accessible if there were an event that~would require its use. Resolution: Modify the system to make it more usable, reliable, and maintainable. Closure Requirement: ECNs A-47111, A-5773, R-0285, R-0317, R-0361, R-0465, R-0559, R-0562, R-0732, R-0337, A-5735. Completion Date: Prior to plant restart. 4C.41.3 Problem 3 Review of open work requests - priority 1. Tracking Number: 26.0471
Description:
Open work requests were reviewed for work items that must be completed prior to restart. Work requests 110635, 116288, 116291, 111534, 111535, and 110516 must be completed prior to plant restart. 4C.41-1
l
)
Resolution: Complete work and close out work requests. Closure Requirement: Work requests 110635, 116288, 116291, 111534, 111535, and 110516. Completion Date: Prior to plant restart. 4C.41.4 Problem 4 Review of open NCRs - priority 1. Tracking Number: 26.0474
Description:
Open nonconforming reports were reviewed for items that must be completed prior to restart. NCR S-5609 must be closed out prior to restart. Resolution: Complete work and close out the NCR. Closure Requirement: NCR S-5609. : 1 Completion Date: Prior to PASS acceptance test. 4C.41.5 Problem 5 PASS needs to be modified to reduce radiation levels. Tracking Number: 26.0978
Description:
The postulated post-accident dose rates at the PASS control panel and intrinsic germanium detector are high. The PASS needs to be modified to reduce the radiation levels in the area. Resolution; 1) reroute the coolant sample line (70057-3/8"-CA) and drain line (70057-5/8"-CA) to eliminate the runs of piping beneath the PASS control panel.
- 2) Provide local shielding for the pre-SCAS coolant sample line.
Closure Requirement: ECNs to perform modifications. Completion Date: Prior to PASS acceptance tert. 4C.41.6 Problem 6 Need a contract for off-site laboratory support. Tracking Number: 26.0980 i 4C.41-2
Description:
The contract for off-site laboratory support for PASS
/ will expire at the end of 1986. The contract needs to
(' be renewed or another contract needs to be in place prior to plant restart. Resolution: Renew the contract or obtain a new contract for off-site laboratory support for post-accident sampling. Closure Requirement: A new contract in place. Completion Date: Prior to plant restart. 4C.41.7 Problem 7 R-15713 does not have sufficient range. Tracking Number: 26.0981
Description:
In its present configuration, the radiation monitor , adjacent to the PASS sample inlet line does not provide an adequate range to determine the required dilution factors for PASS. Resolution: Modify the radiation monitor.
,_ Closure Requirement: ECN to perform modification.
I V Completion Date: Prior to PASS acceptance test. 4C.41.8 Problem 8 NRC open inspection items should be closed. Tracking Number: 26.0982
Description:
NRC open inspection items for PASS should be closed prior to restart. Resolution: Address and close NRC open inspection items for PASS. Closure Requirement: No open inspection items. Completion Date: Prior to plant restart. 4C.41-3
l 4C.42 Emergency Diesel Generator System (EGS) O 4C .42.1 Problem 1 Diesel Generator Annunciation. Tracking Number: 20.0105
Description:
The emergency diesel generator was found to be inoperable as a result of relay operation causing a ' lockout. The relay operation was improperly annunciated and indicated. Resolution: Modify'the emergency diesel generator annunciator system. Closure Requirement: ECN's A-3881 Sub A and R-1000. Completion Date: Prior to G.M. Diesel functional testing.
/ \
4C.42.2 ' Problem 2 The "not ready for auto start" annunciator actuates for normal diesel starts. Tracking Number: 26.0182
Description:
The Diesel Gen A(B) start inoperable / loss of remote control annunciator actuates during routine start of the diesel. Resolution: Eliminate the nulsance alarm by circuit modifIcatIcn. Closure Requirement: Resolve problem and test satisfactorily. Completion Date: Prior to G.M. Diesel functional testing. May be tested during functional testing. 4C.42.3 Problem 3 Review of open diesel generator work requests (prior to 07/12/86). t Tracking Number: 21.0180 0
- 4C.42-1
v 3, , (: '
Description:
Work Request Number Description Engine 117235 Do SP 206.01A. A 117236 Do SP 206.018. B 115496 Provide NPS support to clean the A air box and lube oli sump. 116667 Drain lube oil and coolant from A tr4gine into clean 55 gallon drums, save for reuse. 117224 Replace turbocharger. A 117225 Replace stubshaft and bracket. A 117045 Replace turbocharger. (117226 Volded) B 117046 Replace stubshaft and bracket. 8 (117227 Volded) 112938 4KV splices appear to be deteriorating A f in H20GA cabinets, and various types of hardware have been used for terminating lugs. 116165 Rework Panel H2DGA-L2 wiring to allow A access to terminal boards 1 and 21 per NCR 4640 Rev. I disposition. 116180 Add a Hut Raychem K1t to the CT count A box end of cable 1 PIA 0BP. s' 116728 Insulation on start leads damaged by A
,, j clamping block.
T 116768 Manufacture new bus for "A" Diesel A 2* Generator PT's as per samples. 117249 Rework lugs in accordance with NCR A disposition. , 116178 Implement troubleshooting action plan A a s , for diesel generator light bulb.
'\\
\
116181 Implement troubleshooting action plan for diesel generator light bulbs on A ! quarantined annunciator panel.
'l- 116617 Support flush and vent of Woodward A Governor during/after SP 206.01A.
116781 Verify proper CT connection per EM.188-192. A 105591 Attach thermocouple wells. 116614 24" exhaust duct flange to transition piece A is lacking 3" of weld. 116615 Inspect upper and lower middle brace welds. B ! (Ref. NCR 56B1) 117439 Valve leaks during operation. 118076 Install thermocouple wells. 8 118118 Clean accessible areas of turbo exhaust duct. 117315 Replace vacuum indicator; it sticks. 117147 011 leaking by lube oil dipstick B 116172 Prefab a stainless steel shroud / air baffle A 116671 Provide mechanical support for generator B removal / installation.
~
4C.42-2
H
* , Resolution: Implement the above listed work requests as necessary for
(' closure. M , Closure Requirement: Close the above listed work requests after work is complete. a/. Completion Date: Prior to G.M. Diesel functional testing. 1 4C.42.4 5Y Problem 4 The diesel speed tachometer relays are obsolete and unreliable. Tracking Number: 26.0360 De nription: The existing Dynalco speed tachometer relays have been obsolete for some time and have exhibited setpoint drift. Resolution: Modify the emergency diesel generator control system by upgrading the speed tachometer relays. Closure Requirement: ECN R-1108. Completion Date: Prior to G.M. Diesel functional testing. 4C.42.5 Problem 5 The on-site lube oil supply has not been determined sufficient to maintain diesel generator operability. Tracking Number: 26.0374 Des'cription: The minimum on site lube oil supply has not been determined sufficient to maintain the diesel generators operating for seven days. Resolution: Provide an administratively controlled seven day tube oil supply. Closure Requirement: Completed. Completion Date: Complete. 4C.42.6 Problem 6 The load sequencer may be bypassed following a demand for Diesel Generator operation. Tracking Number: 20.0104 I 4C.42-3
== Description:== Under certain situations, the load sequencer to the diesel generator may be bypassed. Resolution: Modify sequencing circuitry. Revise DG operating procedures to dissallow load sequence bypass under any normal conditions. Closure Requirement: CKT modification completed. OP.A31 revised. Completion Date: Prior to LOSP and ESF actuation (LOSP) or initial plant Heat Up (HTUP), if testing is required. 4C.42.7 Problem 7 Review of open Nonconformance Reports (NCRs). f Tracking Number: 26.0680
== Description:== NCR NUMBER DESCRIPTION - ( S-5459, Rev. O Incorrect output fuses (F1-F2) installed in the GEA Diesel Generator regulator. Work Request #112719. (Complete) S-5528, Rev. O Air Compressor C-891-B motor bearings were replaced without a QC inspection. Work Request 114849. S-5529, Rev. O Air Compressor C-889-8 motor bearings were replaced without a QC inspection. Work ficquest 114848. S-5714, Rev. 0 #1 Idler Gear Bushing indicates abnormal wear. Work Request 117224 (Complete). ECN #R-0770A. S-5727, Rev. 1 Use of incorrect bearing grease is causing grease decomposition, shroud damage and a fire hazard. Work Request 117244 (Complete), 115593 (Complete). ECN #R-0890. Resolution: Implement NCR disposition as necessary for closure. Closure Requirement: Verify NCR disposition actions are complete. Completion Date: Prior to G.H. Diesel functional testing. O 4C.42-4
l 4C.42.8 (M (,v) Problem 8 Review of open Diesel Generator Engineering Change l Notices (ECNs). Tracking Number: 26.0682
Description:
ECN NUMBER DESCRIPTION R-0770A, Rev. O Upgrade the existing turbochargers and the associated gears to better withstand light load operation. Work Order 108527. MOD #527. R-07708, Rev. 0 Upgrade the No. 1 Idler Gear and Stubshaft assembly to the current design for improved strength and reliability. Work Order 108527. MOD
#527.
R-0890, Rev. 0 1. Replace space heater shroud, damaged from exposure to hot liquified grease, with stainless steel. NCR S-5727.
- 2. Replace damaged terminal blocks in panel H2DGA CT connection box and L2 cabinet. NCR 5-5730. Work Order 592006.
b ( R-1000, Rev. 0 1. Provide a control room alarm on loss of diesel generator output breaker DC control power.
- 2. Indicating lights will be added to the local diesel generator panels H2DGA and B to indicate the following conditions:
- a. Loss of DC control power to diesel generator output breakers S4A08 and S4Bil.
- b. 486 lockout of diesel generator output breakers S4A08 and S4Bil.
- c. 486 lockout of start up transformer #2 breakers S4A01 and S4B04.
- 3. Replace existing and additional indicating lights on the diesel engine and generator control panels related to "not ready for auto start" with yellow lenses. Restart
#1.151. MOD #121.
O 4C.42-5
l R-0415, Sub A, Install a time delay relay in the diesel Rev. O generator output breaker control circuit to allow the NS bus voltage to decay (following load stripping) prior to output breaker closure. Resolution: Impiement the above Iisted modifications. Closure Requirement: Completion of the above listed modifications. Completion Date: Prior to G.M. Diesel functional testing. 4C.42.9 Problem 9: The Diesel Generators could be overloaded when paralleled to offsite power. Tracking Number: 26.0956
Description:
When paralleling a Diesel Generator to offsite power, there are two limits related to KVARS (reactive load). The design KVAR limit is based on a 0.8 lagging Power Factor, whereas, the field amp (excitation current) limit is based on rated voltage. Exceeding either limit is considered an overload condition. Resolution: Prior to restart, review OP A.31 and SP.206.03A(B) to: limit Power Factor to 0.85 lagging; increase monitoring of local and remote parameters during diesel generator operation; and provide a capability curve which Indicates the generator operating envelope. Closure Requirement: Revised OP A.31 and SP 206.03A(B) Completion: Prior to restart. O 4C.42-6
l 4C.43 Seal iniection and Make-Up (SIM) 4C.43.1 Problem 1 Operators should be trained to achieve and maintain an open suction path for make-up pumps. Tracking Number: 15.0059-1
Description:
Training is required to reinforce the need for maintaining an open suction and discharge flow path for make-up pumps. Resolution: Provide operator training to reinforce the need for maintaining open suction and discharge flow paths to make-up pumps. Closure Requirement: Documentation from Training. Completion Date: Initial plant heat up. 4C.43.2 Problem 2 Erroneous indications from high-pressure injection flow indicator at low flow rates. Tracking Number: 15.0113
Description:
Revise Emergency Procedure Rule No. 2 for high-pressure injection flow control to incorporate a caution that flow indication is inaccurate at low flows. Resolution: Revise Emergency Operating Procedures to include this caution. Closure Requirement: Issuance of rewritten E0Ps. Completion Date: Reactor start up. 4C.43.3 Problem 3 Throttling of high-prestwre injection valves at low flow rates is not the most effective flow control method. (<100 gpe) Tracking Number: 15.0114 2
Description:
Investigate whether, at low high-pressure injection flow rates, one or more high-pressure injection valves should be closed rather than throttling below 100 gpm per nozzle. Modify emergency procedures accordingly. 4C.43-1
Resolution: Revise Emergency Operating Procedures to require sequential closing of HPl valves when throttling flow. ' Closure Requirement: Issuance of rewritten E0Ps Completion Date: Reactor start up. 4C.43.4 Problem 4 There is no procedural verification that the make-up tank outlet valve is open before the make-up pumps are started. Tracking Number: 15.0153
== Description:== The emergency operating procedures do not require that the make-up tank isolation valve be open before miniflow (recirculation) is re-established when the high-pressure injection mode is complete. Failure to follow this procedure could result in overfilling the make-up tank. Resolution: Revise Emergency Operating Procedures Rule 2 to require opening SFV-23508 before re-establishing HPl miniflow. Closure Requirement: Issuance of rewritten E0Ps. Completion Date: Reactor start up. 4C.43.5 Problem 5 Open work requests. Tracking Number: 26.0225
== Description:== A review of open work requests was conducted to determine which would require closure prior to plant start up. Work Request Number Problem Description VALVES 112740 Repair SIM-040 body to bonnet and packing leak ("A" HPI inside RB isolation stop check). 11306B Replace worn components in LV-21503, pressurizer level control valve. O 4C.43-2
~
VALVES (Continued)
.(' O) 113151 Fabricate stainless steel packing spacer rings for LV-21503 to replace supplied aluminum ones. This will be investigated for design change implications. ,
112461 I&C to help restore LV-21503 actuator's control air and circuitry. 110590 Repair SFV-23604 packing leak (make-up isolation). 113763 Adjust SFV-23604 operator so valve will pass LLRT. , 111519 Test SFV-23616 per SP203.03 (seal inject filter isolation). 97905 Repair HV-23801 seat leakage, HPI loop B motor operated stop-check (NCR S-4742). 97909 Adjust HV-23801 operator so valve will pass LLRT. 112743 Repair packing leak on HV-23801 114500 E.M. support M.M. to reset HV-23801 limit switches. 97906 Repair HV-23502 seat leakage, alternate PZR spray line isolation (NCR S-4742). 97908 Adjust HV-23802 operator so valve will pass LLRT. 112744 Repair packing leak on HV-23802. 105894 Repair SFV-23810 packing leak, HPI to "B" loop isolation. , 105897 Repair SFV-23811 packing and body to bonnet leak, HPI to "A" loop isolation. 113762 Adjust SFV-23811 operator so valve will pass LLRT. , 111132 Test SFV-23811 per SP203.02A 108284 Repair SFV-23812 packing and body to bonnet leaks, HPI i to "B" loop isolation. 114183 Repair packing leak on high root for FT-23603, flow transmitter upstream of LV-21503. 114184 Repair packing leak on low root for FT-23603. 115049 Reposition cocked flange downstream of SIM-538, "B" loop HPI (inside RB) line vent. 4C.43-3 i
- . , -__-..__,,,_,.,.,,._.m.,.._.,,,m._,,-.__.-_ ., _ ,,,_, _ , - .- -_,.,.. __,, ,,__.,,_,_.,..m_.- ,, . , , _ , , _ . - _ . . . , , . -
MAKE-UP PUMP P-236 114821 P-236 casing weld repair, machining, and reassembly. (NCR S-5241) 110519 P-236 gear drive disassembly, inspection, and rework. (NCR S-5241). 109376 I & C support P-236 instrument connections and interference removal. (NCR S-5241). 109375 EH. support P-236 electrical connections and bracket removal. (NCR S-5241). 115512 Install Raychem tubing on P-236 motor leads (NCR S-5573). 113022 Decontaminate P-236 motor stator. 115376 Machine Four P-236 casing studs. Remove P-236 motor and replace stator. 114099 I&C support P-236 motor removal and reinstallation. 113120 Remove unusable P-236 parts, clear disassembly tent for decontamination, prepare for welding and machining repairs, and move pump casing to tent. 113122 Clean P-236 pump studs and nuts, and inspect damage. 108877 Repair P-236 "STOP" BLPB on H2SFB. 114994 Return computer point T-071 to monitor, when work on P-236 is complete. OTHER 113116 Flush line 23620-2 1/2" - CA from drain at SIM-071 to drain at SIM-491 (make-up line through LV-21503). 115402 Test snubber ISW-23626-3A; repair, retest as necessary. 115471 Test snubber ISW-23822-7B; repair, retest as necessary. 115459 Remove 1SW-23822-78, reinstall after testing. 107651 Replace missing support bracket U-bolt near SIM-070. 110916 Replace existing pipe support on line 23829-2"-CA near SIM-086 with one that is project Class 1. O 4C.43-4
OTHER (Continued)
^%
Q 111700 Install missing pipe strap bolt on pipe support 4G-23635-3, near SIM-109. (NCR S-5227) Collect operator / valve data per IE Bulletin 85-03 for: 115550 SFV-23809-L, HPI to loop "A"-isolation 115551 SFV-23810-L, HPI to loop "B" isolation 115552 SFV-23811-L, HPI to loop "A" isolation 115553 SFV-23812-L, HPI to loop "B" isolation 115559 SFV-23616-L, seal to RCPs 115572 SFV-23604-L, make up to RCS 1 solation 115573 SFV-23645-L, HPI pumps recirc isolation 115574 SFV-23646-L, HPI pumps recirc isolation Resolution: Complete and close out Priority 1 work requests. Closure Requirement: Closure of W/R listed. Completion Date: Initial plant heat up. 4C.43.6 Problem 6 Open nonconformance reports. Tracking Number: 26.0409
Description:
A review of open nonconformance reports was conducted to determine which would require closure prior to plant start up. Resolution: Complete and close out priority 1 NCRs. Closure Requirement: S-5217 Bent SIM Pipe Support (lG-21006-2) S-5445 SFV-23645 Torque Switch Cracked S-5503 SFV-23645 Torque Switch Discrepant S-5697 Oversize Hearing Ring Grooves, P-236 S-5716 P-236 Dowels / Holes Mismatch S-5918 P-236 Unapproved Supplier Completion Date: Initial plant heat up. 4C.43-5
4C.43.7 Problem 7 Open engineering change notices. Tracking Number: 26.0410
== Description:== A review of open Engineering Change Notices was conducted to determine which would require closure prior to start up. Resolution: Complete and close out priority 1 ECNs. Closure Requirement: R-1027 Change vent and tubing connections on P-236. R-0914 85-03 MOV Program. R-0968 85-03 Expanded Program, Safety MOVs. R-1018 Vibration probes for P-236 (the Restart Implementation Manager reviewed this ECN and determined that due to ALARA considerations the work associated with the M/U pump should be completed prior to restart, but that the HPI pump portion should remain as Priority 3. See problem 37; 26.0702. Completion Date: Initial plant heat up. 4C.43.8 O Problem 8 Recurring lube oil leaks on Make-up and HPl pumps. Tracking Number: 26.0103
== Description:== Recurring lube oil leaks on MU/HPI pumps. Resolution: Modify HPl/MU pumps to provide automatic stop of auxiliary lube oil pumps after reaching operating pressure. Closure Requirement: ECN R-1036; NCR S-5134. Completion Date: Reactor start up. 4C.43.9 Problem 9 Qualification of spare stator for make-up pump motor P-263M. Tracking Number: 26.0413 9 4C.43-6
- l
Description:
Since the make-up pump was damaged in the December 26, i 1985, transient, it was decided to replace the old
-(N stator with the new spare stator. (The spare stator i
( ) was procured from Westinghouse in December, 1982 for replacement of the existing stator.) !4o qualification documentation was found associated with the purchase l order for this procurement. ) l Resolution: Provide qualification for make-up pump stator. 1 Closure Requirement: Documentation of qualification from QA. Completion Date: Reactor start up. 4C.43.10 Problem 10 The repair schedule for P-236 is not scheduled. Tracking Number: 16.00178
Description:
Licensee to provide schedule for repair and replacement of the make-up pump. This is a RRR8 priority 1 item. Resolution: Close out this item. Closure Requirenant: P-236 Repair close-out documentation. p
\ Completion Date: Reactor start up.
4 t I l 4C.43-7
4C.44 Carbon Dioxide System (CO2) n 4C.44.1 . Problem 1 Priority one Open CO2 Work Requests. (Prior to ! 9-2-86) Tracking Number: 26.0603
Description:
Plant maintenance daily work listing was reviewed for open work requests that must be completed prior to start up.
- 1. HR 115541 - retest NSEB CO 2 , zone 76. (NCR S-5317)
- 2. HR 118598 - replace flex on cooling fan motor.
- 3. HR 123111 - replace cable to compressor.
- 4. HR 110765 - repair leak in CO2 line leading into zone 77.
Resolution: Nuclear Plant Departments to complete open work requests. Closure Requirement: Complete HR 115541, 118598, 123111 & 110765 Completion Date: Prior to restart (criticality). ; N 4C.44-1
. 4C.45 Fire Protection System (FPS) h 4C.45.1 Problem 1 Spurious fire alarms received during the 12/26/85 event.
Tracking Number: 26.0537
Description:
Early in the 12/26/85 event, the Control Room received numerous spurious fire alarms. Operator action was required to silence the alarms and inspect the areas. The Technical Support Center pre-action sprinkler system was charged due to a momentary interruption of power. Resolution: Modify the fire alarm power supplies to prevent spurious alarms and inadvertent actuation. Closure Requirement: ECN-1079, ECNA-4382, NCR S-2097, HR R-ll42, A-4392. Completion Date: Prior to reactor start up (R"SU). n v 4C.45.2 Problem 2 Fire Zone 104 (TSC) has delayed backup power supply. Tracking Number: 22.0026
Description:
Fire Zone 104 has delayed backup power supply. Resolution: Modify Zone 104 Fire System backup power supply. Closure Requirement: ECN-1079, ECN-R-1078 l Completion Date: Prior to reactor start up (RXSU). 4C.45.3 Problem 3 Manual restart of Auxiliary Building ventilation following shutdown from fire protection interlocks is not possible. Tracking Number: 26.0541 J 4C.45-1
Description:
Auxiliary Building ventilation and other power block ventilation exhaust fans are shut down by fire protection system interlocks. The fans cannot be i manually restarted to clear the smoke because of the ' interlock. A Technical Specification violation results when the Auxiliary Building loses its negative pressure and an unmonitored radioactive release begins. Resolution: Modify the ventilation fan controls to provide bypass of fire protection interlocks. Revise procedures and provide operator training. i Closure Requirement: ECN-R-1142 Revise procedures. Document operator training. Completion Date: Prior to reactor start up (RXSU). 4C.45.4 Problem 4 A water leakage path exists from TSC to east 480 volt switchgear room. Tracking Number: 26.0545
Description:
Install permanent drain for the Technical Support Center. Seal off leakage path to east 480V switchgear room. Resolution: Install permanent drain for the Technical Support Center. Seal off leakage path to east 480V switchgear room. Closure Requirement: ECN R-0197 Completion Date: Prior to reactor start up (RXSU). 4C.45.5 Problem 5 Open work request on seismic gap fire seal. Tracking Number: 26.0565
Description:
Open Work Request 111927 provide a fire seal for the l Seismic Gap. l l Resolution: Complete the Work Request to seal the Seismic Gap. 1 Closure Requirement: HR 111927 l Completion Date: Prior to reactor start up. O 4C.45-2
4C.45.6 i Q Problem 6 Complete NSEB Fire Suppression Cross Zone Detection and NSEB NVAC Damper Circuit Separation Tracking Number: 26.0574
Description:
Complete Mod 136, ECN R-0458 NSEB Fire Suppression Cross Zone. Detection and Mod 138, ECN R-1078 NSEB HVAC Damper Circuit Separation. Resolution: Install Mods 136 and 138. Closure Requirement: ECN R-0458 and ECN R-1078. Completion Date: Prior to reactor start up (RXSU). 4 4C.45.7 Probiem 7 Emergency 1ighting and communications for NSEB. Tracking Number: 26.0578
Description:
Emergency lighting, and communications should be installed in the NSEB. Emergency lighting is a safety consideration. Resolution: Install additional emergency lighting and communications to the NSEB. Closure Requirement: ECN A-3867(E), WR-104160 Completion Date: Prior to reactor start up. 4C.45.8 Problem 8 Appendix R compliance for Auxiliary Feedwater Pumps, i Tracking Number: 26.0580
Description:
The two Auxiliary Feedwater Pumps (P-318, P-319) located within the missile cage of the Nuclear Service yard area do not meet the separation requirements of the 10CFR50, Appendix R III. G.2. In order to satisfy this requirement, the district has submitted an exemption request which calls for the addition of a deluge sprinkler system above the two pumps. Resolution: Install deluge sprinkler system in the AFW pump area lO 4C.45-3
- .- _-___ _ .~.__ _.__ _ _ _ - _ . _ . . _ _ _ . _ _ . . _ _ _ _ _ _ _ - . _ . _ . -
Closure Requirement: ECN A-5530, NCR S-5331, S-5332, S-5373, S-5664, HR { 105433, 105434, 105436, 115226 Completion Date: Prior to reactor start up (RXSU). 4C.45.10 Problem 10 Fire Protection Systems overdue for surveillance. Tracking Number: 26.0587
Description:
Revise SG.104 using current Technical Specifications and identifying all Surveillance requirements. Revise acceptance criteria of SP.201.038 to include pressure and RPM criteria for diesel driven fire pumps. Open item 85-36-01: Ensure approval of revised fire protection administrative procedure. Resolution: Revise SG.104 and SP.201.03B. Closure Requirement: SG.104, SP.201.038 Completion Date: Prior to reactor start up (RXSU). 4C.45.11 Problem 11 Control circuitry and equipment to support operability of NSEB Ventilation System is unreliable. Tracking Number: 26.0574
Description:
Control circuitry and equipment to support operability of NSEB Ventilation System is unreliable. Resolution: Perform ECN R-1078 9 4C.45-4
Closure Requirement: Completed ECN O] ( Completion Date: Prior to plant heat up. 4C.45.12 Problem 12 NSEB Cable Shaft Fire Protection is inadequate. Tracking Number: 22.0086
Description:
NSEB Cable Shaft Fire Protection is inadequate. Resolution: Perform ECN R-0485 Closure Requirement: Completed ECN Completion Date: Prior to plant heat up. ' Closure of this item satisfies QTS 15.0193, QTS 26.0545, open item 86-18-07. 4C.45.13 Problem 13 Fire Barrier Scope should be completed in accordance with Appendix R. C Tracking Number: 26.0565
Description:
Fire Barrier Scope should be completed in accordance with Appendix R. Resolution: Complete ECN R-0838 , Closure Requirement: Completed ECN Completion Date: Prior to plant heat up. Closure of this item satisfies QTS 26.0565. 4C.45.14 Problem 14 Fire Protection Systems in new DG building, Aux. Buliding, and Tank Fara should be upgraded. Tracking Number: 26.0580
Description:
Fire Protection Systems in new DG building, Aux. Building, and Tank Farm should be upgraded. O 4C.45-5
Resolution: 1) ECN R-0767
- 2) ECN R-0527
- 3) ECN R-1185 Closure Requirement: Complete ECNs Completion Date: Prior to plant heat up.
O O 4C.45-6
4C.46 Main Condensate and Makeup System (MCM) 4C.46.1 Problem 1 Condensate pump fuse is wrong style. Tracking Number: 21.0171A
Description:
A substitute fuse has been installed in the main condensate pump starting circuitry. . Resolution: Change the fuse to the correct style. Closure Requirement: Fuse changes - QC inspection results. Completion Date: Prior to drawing condenser vacuum. 4C.46.2 Problem 2 Open Work Requests on MCM and AES - Priority 1.
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Tracking Number: 26.0909
Description:
7 Work Requests on Condensate Pump P-351 B, 7 WRs are on non-functioning computer points, 2 hrs are to repair h V instruments, and 13 are on valves. Resolution: Complete open WRs. ~1 Closure Requirement: NCR S-5673; HR 109415,109486,116319,117900,116579, 120032, 120033, 112781, 112885, 114206, 115141, 123148, 1231C 2, 123231, 113657, 118642, 097986, 115391, 115392, l 123047, 123048, 117278, 117488, 117265, 117080, 119506, ; 102797, 117557. l Completion Date: Prior to drawing condenser vacuum. j l 4C.46.3 j Problem 3 Condensate Polisher Problems Tracking Number: 26.0907
Description:
PSV-34508 is missing; PSV-34508 is set at 501 1 psig. Level drain valve and the air, water drain, and recycle bottom tank valves leak by when condensate pressure is applied.
- o l i
4C.46-1
Resolution: Install PSV-34508 per WR 82211. Replace existing leaking valves per ECN A-4968. Adjust setpoint of PC-34509 to 301 1 psig (per Rev. 9 of AP.152) Closure Requirement: HR 82211, ECN A-4968, NCR S-6079, EAR #PL 86-609, AP.152, NCR S-5657. Completion Date: Prior to drawing condenser vacuum. 4C.46.4 Problem 4 Drain Line off Bottom of CST locks seismic supports. Tracking Number: 21.0172B
Description:
Drain Line off bottom of CST lacks seismic supports. Resolution: Install pipe supports IAW ECN R-0147 Closure Requirement: ECN R-0147 Completion Date: Prior to drawing condenser vacuum. O O 4C.46-2
4C.47 Plant Buildings and Structures (PBS) 4C.47.1 Problem 1 Review of open NCRs - Priority 1 Tracking Number: 26.1009
Description:
The fcilowing Open Nonconforming Reports will be dispositioned prior to restart. NCR S-5075 - local leak rate tests RSP-55 excessive leakage was noted. NCR S-5238 - local leak rate tests on RSP-34 pressure could not be maintained above 52 psig. Resolution: Repair valves RSP-34 and 55 to meet acceptable leakage rates. Closure Requirement: Completion of work requests and closure of NCR's. Completion Date: Unit S/U turnover to dispatch. 4 40.47.2 Proolem 2 Inadequate access to locked areas. Tracking Number: 15.0016
Description:
Inadequate access to locked areas of the Auxiliary Building to Health Physics and Operators might delay the response to accident conditions. Resolution: Change all locks on the -20 and -47 to allow opening with a single key. Closure Requirement: Hork Request completion. Completion Date: Unit S/U turnover to dispatch. ! 4C.47.3 Problem 3 Vital area penetration does not meet the 10CFR requirements. Tracking Number: 19.0026 O ! 4C.47-1
Description:
The fire hose cabinet next to Door AU121 in the Auxiliary Building has been modified to meet Appendix R specification. Due to the construction of the cabinet, the penetration did not meet the 10CFR requirements for a vital area boundary. Resolution: Modify vital Ares 12 wall behind fire hose station 10, adjacent to Door AU121. Closure Requirement: ECN closed Completion Date: Unit S/U turnover to dispatch. 4C.47.4 Problem 4 Review of open WR - Priority 1. Tracking Number: 26.1007
Description:
The Plant Maintenance Daily Work Listing was reviewed for open work items (prior to 11-04-86) that must be completed prior to restart. Resolution: Complete work and close out work requests prior to restart. Closure. ,1rement: HR110015 Completion Date: Start ILRT 4C.47.5 Problem 5: Review NCR IE Information Notice No. 85-71 for applicability to Rancho Seco. Tracking Number: 27.0032
Description:
NRC IE Information Notice No. 85-71 (Containment Integrated Leak Rate Tests) shall be reviewed for applicability to Rancho Seco. Resolution: NRC IE Inform **lon Notice No. 85-71 will be reviewed and applied to the integrated leak rate test prior to the performan, of the ILRT. Closure Requirement: Engineering Re ,. nse and incorporation of any required changes to ILRT procedures. Completion Date: Prior to ILRT performance. O 4C.47-2
l l I 1 4C.48.1 Control Room /TSC Essential HVAC (MVS) f% i k 4C.48.1 Problem 1 Test agent injection manifolds necessary for in-place leak testing have not been installed and tested. Tracking Number: 26.0016
Description:
HEPA and carbon filter banks in the Essential Filtration Units must be tested periodically for leaks to satisfy Technical Specification 4.10 requirements. This testing is accomplished by injecting a test agent into the airstream ahead of each filter bank. Injection manifolds are necessary to ensure all regions of the filter banks are uniformly challenged. The details of this are included in ANSI N S10, Testing of Nuclear Air-Cleaning Systems. Test manifolds have not been installed in the Essential Air Filtration Units. Resolution: Install test agent injection and sampling manifolds. Closure Requirement: ECN R 0788, NCR S 4761, STP 1063A and B Completion Date: Prior to performance of SP211.018 and SP211.01E (Refueling Interval Surveillance Test). 4C.48.2 Problem 2 Excessive noise in the Control Room caused by operation of the Essential HVAC System. Tracking Number: 19.0010.A
Description:
Operation of each train of essential HVAC equipment produces undesirable sound levels in the Control Room and impairs communication within the Control Room and .: while using the telephone. When both trains of equipment operate simultaneously, as occurs during i automatic actuation of the essential system, the sound levels increase and have a severe impact on 7 communication. Resolution: Perform a special EST to obtain data on the affect of reduced air flow on CR/TSC noise levels and system j cooling capability. Evaluate data and implement , resolution per the SSR. l Closure Requirement: Completion of STP 1010 ECN 1402, and RT-HVS-010 Air ! Balance. O 4C.48-1
Completion Date: Prior to initial heat up. 4C.48.3 O , Problem 3 Control Roon Normal HVAC (Ficw Switch) is in the wrong flow path. Tracking Number: 26.0017
Description:
A paddle-type low flow switch was installed in the cold air supp'y duct to shut off the normal air handler upon actuation of the Essential HVAC System. During the colder periods of the year there is a smaller demand for cooling in the Control Room, such that the air flow I through the cold duct is decreased. This air reduction was enough to cause the flow switch to trip and shutdown the normal air handler, resulting in a loss of normal cooling to the Control Room. After a short time, the control Room heats up and causes the Essential HVAC System to automatically actuate upon Control Room high - temperature. } Resolution: Remove the low flow switch from the Normal System cold air supply duct and reinstall in the Normal System return air duct. Closure Requirement: ECN R 0314 Completion Date: Prior to start of CR/TSC Essential Air System functional testing (about February 15, 1987). 4C.48.4 Problem 4 The provisions exists for testing of the Control Room /TSC Essential HVAC (Pressurization Measurement Access for TSC) Tracking Number: 26.0001
Description:
Technical Specification 4.10 requires that the Control Room /TSC Essential HVAC System have capability to pressurize the Control Room and TSC to 0.125 ING (min.); however, no pressure tap exists to measure the pressure in the TSC. Pressure readings are presently taken by putting a tube under the door into the TSC. Resolution: Install pressure taps per ECN R-0938. Closure Requirement: ECN R 0938 complete Completion Date: Prior to Control Room habitability testing (to be performed by NRC). 4C.48-2
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4 C.48.5 p
\ Problem 5 Power supplies to Essential HVAC Units do not conform to normal engineering practice for providing reliable power to the controls and instruments.
Tracking Number: 26.0153
Description:
The control power supply for the filtration units SF-A-7A and SF-A-78 (power supply units and Raymond actuators) is supplied from 120V distribution panels which could have steady state voltage availability below that which is recommended for operation of the equipment. The Class 2 instruments of the flitration and air handling units share the same power supplies with the lighting and receptacles of the units. This does not follow established practices for providing power to instruments. Resolution: Control power shall be relocated from the 120V distribution panels S1A3 and S1B3 to the distribution panels $1 A2-1 and S182-1.
- liluminating lights shall be provided with a new control power supply.
I O 4C.48.6 Problem 6 CR/TSC air flow transmitter current loop impedance i causes transmitter output to be out of tolerance low between 75 and 100% of full range. Tracking Number: 26.0481
Description:
A current loop is used to generate the flow transmitter output signal. Excess impedance in the current loop prevents the development of full output voltage, 1 thereby limiting the transmitter at the hi end of the output range. Resolution: Recove existing flow switches and computer points from input loop and relocate per ECN R 1091. Closure Requirement: ECN R 1091,NCR S 5863. , 4 Completion Date: Prior to beginning post modification testing of CR/TSC i Essential Air Syste" (About February 15, 1987). 4C.48-3
4C.48.7 Problem 7 The quantity of refrigerant in the CR/TSC Essential Air refrigeration system is difficult to determine during maintenance and system troubleshooting. Tracking Number: 26.0482
Description:
The quantity of refrigerant charge present in the CR/TSC Essential Air refrigerant system is difficult to determine without a receiver site glass. Resolution: Install refrigerant receiver sight glasses. Closure Requirement: ECN R 1160 C Completion Date: Prior to beginning post modification testing of the CR/TSC Essential Air System (about February 15, 1987). 4C.48.8 - Problem 8 Cooling capability of the CR/TSC Essential Air System is marginal during cold weather operation. Tracking Number: 26.0167
Description:
The controls of the refrigeration portion of the HVAC system do not permit reasonable flexibility of the system to respond to all variations in cooling loads with respect to ambient conditions. During low ambient conditions, the system falls to maintain adequate pressure in the suction line resulting in compressor trips. The consequence of this condition is that the essential HVAC system cannot be relied upon to operate under all postulated accident modes. Resolution: Modify the controls for the condenser fans to allow cycling of two of the three fans in response to ambient temperatures. ECN R-0769 completed. Closure Requirement: ECN R 0769, STP-1059, STP-1061 Completion Date: Prior to beginning post modification testing of the CR/TSCC Essential Air System (about February 15, 1987). 40.48.9 Problem 9 Control Room /TSC Essential HVAC System (Refrigeration Problems). Tracking Number: 26.0013 4C.48-4
5
Description:
Refrigeration System modifications are necessary to ' O improve performance of the system and to facilitate efficient and effective maintenance. Resolutiu : 1. Modify system controls to provide a system pumpdown capability to return refrigerant to the receiver when the system is shut down. (STP 1059, STP 1060) 2 Modify system cmtrols to provide hot gas bypass controls to prevent feeding hot gas thru an e inactive evaporator coil.
- 3. Design a new P&lD to graphically represent the CR/TSC Essential Air refrigeration system. (PEID 504 Sh 2/3)
- 4. Modify the evaporator control sequence to provide closer incremented control. (STP 1059, STP 1060)
Closure Requirement: ECN R 0904, Drawing M-504 Sheets 2 and 3, STP 1059, STP 1060. Completion Date: Prior to initial plant heat up. 4C.48.10 Problem 10 Work on fire dampers between the Control Room and TSC (. 7
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(currently deferred under M00 111) is not complete. Tracking Number: 19.0011.A j
Description:
Eight existing dampers are to be modified and two new 4 dampers are to be installed. Long term operation with
, compensatory measures is not desirable.
Resolution: Modify eight existing dampers and intuli two new dampers. Closure Requirement: ECN R 0763, ECN R 0764, SP-201.03M, (SP.,1015) Completion Date: Prior to beginning post modification testing of CR/T5C ' Essential Air Systee (about February 15, 1987). 4C.48.11 Problem 11 The CR/TSC Essential Air Conditioning System is subject to spurious actuations caused by high temperature in certain areas of the Control Room. Tracking Number: 26.0018 ) 4C.48-5 l l
\
Description:
Temperature sensors which actuate the CR/TSC Essential Air System in the event of high Control Room temperature (80*F) are located approximately eight feet above the floor in the vicinity of computer equipment. Temperature in this area is somewhat elevated relative to the average temperature in the Control Room. c Spurious system actuation is caused by high temperatures in the vicinity of the sensors rather than excessive average temperature in the Control Room. Resolution: 3alance the Cr/TSE Normal Air System per Maintenance Procedure, RT-HVS-010. Closure Requirement: HRN 120454 Completion Date: Prior to initial plant heat up. 4C.48.12 Problem 12 Need exists to accomplish a Control Room Normal HVAC System Air Flow Balance. Tracking Number: 26.0011
Description:
The air balance of the Control Room Normal HVAC Systeat has not been updated or verified since the original balancing was performed in 1973, despite several modifitations which could affect the balance or the heat load. As a result, the HVAC equipment may not be satisfying the current loading in the Control Room. Resolution: Balance the Control Room Normal HVAC System in accordance with Work Request Number 120454. s Closure Requirement: Work Request 120454 Completion Date: Prior to initial start up. 4C.48,13 Problem 13 The Cr/TSC Essential Air Conditioning System is subject to spurious actuations caused by the Chlorine Gas Detector. Tracking Number: 26.0025
Description:
The chlorine gas sensor contains an electrolytic fluid which must be replenished periodically. Depletion of , this fluid causes the detector to actuate. ' Replenishment of this fluid should be accomplished in conjunction with other preventive maintenance items on a periodic basis. 4C.48-6 a . - __ _ _ _ - _ _ _ _ _ _ _ _ _
l4 Resolution: Perform periodic maintenance on the chlorine gas detector per the Chlorine Gas Detector Periodic Maintenance Procedure. Closure Requirement: Maintenance procedure SP 485 A & B Completion Date: Prior to initial plant heat up. 4C.48.14 i Problem 14 Infiltration through the Cr/TSC Essential Air Handler (AH-A-545A) access door is occurring. Tracking Number: 26.0008
Description:
Air leakage into the Cr/TSC Essential Air Handler "A" has been detected. This is a breach of the Control i Room /TSC pressure boundary resulting in unfiltered air in.iection. Resolution: Replace the access door gasket per Work Request 119806 Closure Requirement: Work Request Number 119806 (RT !!VS-004) Completion Date: Prior to post modification testing. 4C.48.15 Problem 15 Control Room isolation Dampers do not operate properly. l Tracking Number: 26.0004 l
Description:
Contro. mom isolation dampers have experienced difficulty in reaching full-close position and may be leaking. Access doors were not installed near all dampers making maintenance of these dampers very difficult. Ilesolution: 1. Implement a design change to add access doors where necessary for inspecting and maintaining the dampers.
- 2. Review CR/TSC Essential Air isolation dampers to verify installation in accordance with manufacturer's instructions. Contact the supplier to verify that the installation is proper.
- 3. Review and update preventive maintenance requirements for the dampers and their components.
- 4. Implement procedures for periodic maintenance of isolation dampers.
4C.48-7
Closure Requirement: 1. Memo stating additional access doors are not required.
- 2. Completion of inspection, cleaning and lubrication of CR/TSC Essential Air isolation dampers per the following: .
Damper No. HRN Damper No. HRN HV 54705 120275 HV 54717 118883 HV 54706 120276 HV 54718 118884 HV 54707 120253 HV 54719 118885 HV 54708 120254 HV 54720 118898 HV 54709 120251 HV 54721 118899 HV 54710 120252 HV 54722 118900 HV 54715 120249 HV 54723 118901 HV 54716 120250 HV 54724 118902 HV 54733 119450 HV 54725 120245 HV 54734 119449 HV 54726 120246, 120286 HV 54735 120248 HV 54727 115770 HV 54736 120247 HV 54728 118874 HV 54729 118876 HV 54730 119448, 118877 HV 54731 118881, 120278 HV 54732 118882
- 3. Memo deferring maintenance requirement and maintenance frequency.
- 4. Maintenance procedures M 148, M 152 Completion Date: Prior to start of CR/TSC Essential Air System functional testing (about February 15, 1987).
4C.48.16 Problem 16 Control Room Toilet Area Exhaust Fan Isolation damper A (HV-54727) is not positioning properly when actuated and/or damper position displayed by IDADS is not proper. Tracking Number: 26.0002
== Description:== During performance of STP-194 (CR/TSC Essential Air Conditioning Safety Features Operation and Flow Detection Test), HV-54727 was indicated by IDADS to be in both open and closed positions simultaneously when the system was operated in the ventilation (high temperature / radiation) mode. Resolution: Repair or adjust damper as necessary to restore proper function. 4C.48-8
l 1 Closure Requirement: NCR 5611 SP .0084A, SP .0084B, RT HVS-003, RT HVS-005 l Completion Date: Prior to initial plant heat up. 4C.48.17 Problem 17 Shift Supervisor Office Exhaust Fan Isolation damper 8 (HV-54732) is not positioning properly when actuated and/or damper position displayed by IDADS is not proper. Tracking Number: 26.0003
Description:
During performance of STP-194 (CR/TSC Essential Air Conditioning Safety Features Operation and Flow Detection Test), HV-54732 was indicated by IDADS to be in both open and closed positions simultaneously when the system was operated in the ventilation (high temperature / radiation) mode. Resolution: Perform Surveillance Procedure SP 84A, SP 848 Closure Requirement: SP .0084A, SP .00848, RT HVS-003, RT HVS-005 Completion Date: Prior to initial plant heat up. 4C.48.18 Problem 18 Control Room pressure is not maintained at or above 0.125 IWG relative to outside atmosphere by operation of either CR/TSC Essential Air System train. Tracking Number: 26.0006
Description:
During performance of STP-189 (Control Room / Technical Support Center Pressure Test), CR/TSC pressure could not be maintained above 0.125 IWG relative to outside atmosphere by operation of either Cr/TSC Essential Air train. Resolution: Test CR/TSC Essential Air System pressurization capability per STP 189, SP 211.01 A and SP 211.01 D Closure Requirement: STP 189, SP 84A, and SP 848 j Completion Date: Prior to initial plant heat up. i l O ! 4C.48-9
4C.48.19 Problem 19 Operation of the CR/TSC Essential Air Refrigerant Systems is compromised by certain piping and control problems. Tracking Number: 26.0014
== Description:==
- 1. A leak in the CR/TSC Essential Air Train A refrigerant piping is suspected.
- 2. A pressure regulator in the CR/TSC Essential Air Train A refrigerant system is subject to short cycling resultant in continuous surging.
- 3. Evaporator coil superheat should be checked and adjusted.
Resolution: 1. Repair any leaks found in refrigeration piping.
- 2. Investigate surging per WRN 119536 and 119537.
Modify condenser fan controls per ECN R 0769
- 3. Adjust evaporation coil superheat per WR 119815.
Closure Requirement: Closure of HRN 119536,119537,119815, and 119816. Completion of ECN R 0769. Completion Date: Prior to beginning post modification testing of CR/TSC Essential Air System (about February 15, 1987). 4C.48.20 Problem 20 The Control Room does not appear to be uniformly cooled by operation of the Essential Air System. Tracking Number: 26.0012
== Description:== Air distribution in the Control Room with the CR/TSC Essential Air System operating is not believed to be adequate. This problem may be compounded by duct / diffuser modifications for noise reduction. Resolution: Balance the CR/TSC Essentiai Air Distribution System per maintenance procedure. Closure Requirement: Completion of maintenance procedure, RT HVS 010. Completion Date: Prior to initial plant heat up. 4C.48.21 Problem 21 Air flow through the Essential Air Filtration Units may exceed the maximum rates specified in technical Specification 4.10 for several minutes after the systems are actuated. 4C.48-10 l
l Tracking Number: 26.0010
Description:
Booster fan inlet vanes, which control flow through the Essential Filtration Units, are positioned in response to differential pressure developed across the carbon . filter banks. When the units are actuated, there is no l differential pressure; the inlet vanes are opened to the maximum position thereby-allowing an excessive flow rate to be established. This condition persists for several minutes until a stabilized flow rate is established. Resolution: Adjust the Estantial Filtration Unit flow controls to l reduce the time nocessary to establish stabilized flow. ! Closure Requirement: Completion of work request. Completion Date: Prior to initial plant heat Lp. 4C.48.22 Problem 22 There is no approved procedure for calibration of the CR/TSC Essential Filtration Unit Flow Transmitter. Tracking Number: 26.0009 O
Description:
Differential pressure across the carbon filter banks is the parameter measured to control flow thru the Essential Filtration Units. The flow characteristics are subject to change when the carbon is changed, necessitating recalibration of the flow transmitter to assure the proper flow rate is maintained and proper indication of flow rate is indicated in IDADS. Resolution: Write a procedure for calibration of Essential Unit , flow transmitters. ; Closure Requirement: New calibration procedure l Completion Date: Prior to beginning post modification testing of CR/TSC Essential Air System (about February 15, 1987). 4C.58.23 Problem 23 CR/TSC Essential Filtration Unit A Flow Transmitter (FT-54701) and/or related IDADS components:
- 1. Operate intermittently and/or provide intermittent flow indication on the IDADS monitor.
O 4C.48-ll
- _ _ _ _ _ _ _ _ _ _ __ - - . - . _ _ _ _ . ~ . . _ _ . . . _ _ . _ _ _ _ _ _ . _ _ _ - . . _ _ _ _ . _ _
- 2. Control flow at a rate which is in excess of the Technical Specification 4.10 flow limit and/or indicate a flow rate which is in excess of the Technical Specification 4.10 flow limit.
Tracking Number: 26.0005
Description:
Flow control and indication of the filter unit is accomplished by a Foxboro Transmitter which monitors the pressure differential across the carbon filters and converts this into an electrical signal, which controls the booter fan inlet vanes in the unit and inputs to the IDADS computer. The computer then processes this signal in an equation and displays the flow rate in cfm. During performance of STP-194, IDADS Indication of flow through the Essential Filtration Unit A (SF-A-7A) was intermittent and indicated flow rate exceeded Technical Specification 4.10 maximum flow limit. Actual flow rate was determined to be 3010 CFH; within Tech Spec 4.10 Limitations. Resolution: Revise the equation used by IDADS to provide a more accurate IDADS indication of flow rate. Closure Requirement: Memo from Nuclear Plant Manager stating completion - or closed work request.
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Completion Date: Prior to initial plant heat up. 4C.48.24 Problem 24 CR/TSC Essential Filtration Unit B Flow Transmitter (FT-54702) and/or related IDADS components control flow at a rate which is in excess of the Technical Specification 4.10 fIow Iimit and/or indicate a fIow l rate which is in excess of the Technical Specification 4.10 flow limit. Tracking Number: 26.0007
Description:
Flow control and indication of the filter unit is accomplished by a Foxboro Transmitter which monitors the differential pressure across the carbon filter and t converts this into an electrical signal, which controls l the booster fan inlet vanes and inputs to the IDADS computer. The computer then processes this signal into an equation which indicates flow rate in cfm. j During performance of STP-194, IDADS indication of flow l through the Essential Filtration Unit B (SF-A-78) I exceeded Technical Specification 4.10 maximum flow 4C.48-12 l
limit. Actual flow rate was determined to be 3050 CHM; within Tech Spec 4.10 limits. Resolution: Balance the outside and return air flow rates. Revise the equation used by IDADS to provide a more accurate IDADS indication of flow rate. Closure Requirement: WRN 114691 Completion Date: Prior to initial plant heat up. 4C.48.25 Problem 25 Operating Procedure A.14 does not include adequate instructions for operation of the CR/TSC Essential Air System. Tracking Number: 19.0010.0
Description:
Operating Precedure A.14 is deficient in the following areas:
- 1. The difference between Train A and Train B in the present Interim Number 2 configuration is not clear.
- 2. The limitations imposed on the system when in the isolation /stop mode are not delineated.
- 3. Criteria for stopping operation of the individual trains after automatic actuation is not included.
- 4. Interaction with the Nuclear Service Electrical Bus Unioading Schedme is not discussed.
- 5. Instructions for placing system in standby are not complete.
Resolution: Revise Operating Procedure A.14 Closure Requirement: New procedure l Completion Date: Prior to initial plant heat up. AC.48.26 Problem 26 Surveillance Procedures SP .0084A and SP .00848 do not provide a check of all damper limit switch position indications. As a result, damper position and/or position indication problems may be masked. Tracking Number: 26.0023
Description:
There are four damper position IDADS indications for each CR/TSC Essential Air System damper. However, only two of the four positions are recognized on the 4C.48-13 L
pertinent surveillance enclosures. As a consequence, conflicting damper position indications are not detected. Resolution: Revise Surveillance Procedure SP .0084A and SP .00848. Write new Surveillance Procedures, STP 1057 and STP 1060. Closure Requirement: SP 211.01A, SP 211.01D, and new procedure. Completion Date: Prior to initial plant start up. 4C.48.27 Problem 27 Approved procedures for accomptishing non-routine complex refrigeration system maintenance tasks do not exist. Tracking Number: 26.0024
Description:
A Maintenance Procedure describing refrigerant system pumpdown, evacuation, recharging, and compressor change out is needed. The procedure must include a list of required special equipment, materials, etc. Resolution: Write Maintenance Procedures for accomplishing non-routine refrigeration system maintenance tasks. Closure Requirement: M-xxx Freon Charge Removal and Replacement M-xxx Refrigerant Pipe Leak Detection and Repair M-xxx Refrigerant Compressor and Motor Removal and Reinstallation M-xxx Condenser and Hot Gas Bypass Pressure Regulator Adjustment M-xxx Refrigerant Compressor Crankcase Heater Replacement M-xxx Refrigerant Compressor Filter Orter and 011 l Change M-xxx Condenser Fan Removal and Reinstallation M-xxx Evaporator Solenoid Valve and Compressor Unloader Solenoid Coll Replacement Procedure M-xxx Refrigerant Compressor - Motor Coupling Replacement and Check Procedure. t Completion Date: Prior to initial plant heat up. l I 4C.48.28 Problem 28 Operation and control of the Cr/TSC Essential Air System is not understood by Operators. O 4C.48-14
Tracking Number: 26.0021
Description:
Additional training is necessary to enhance Operator knowledge of CR/TSC Essential Air Conditioning System operation. Clarification of backup power sources, . various system operating modes, system availability in the various operating modes, and system interfaces is necessary. Resolution: Provide training to improve operator knowledge of the , Cr/TSC Essential Air Conditioning System. Closure Requirement: Operator training. l Completion Date: Prior to initial plant heat up. I l 4C.48.29 Problem 29 The "hardcast" tape used to seal the Essential Air Handling Units against air infiltration has not been documented as being qualified for long term exposure to expected environmental conditions. Tracking Number: 26.0150 D escription: During the start up testing of the Essential HVAC O System, it was revealed that the air handling units could not meet the leakage criteria of ANSI N509-1980 due to inadequate design and fabrication of the unit housing. "Hardcast" tape was used to seal all of the joints of the housing based on assurance of qualification from the manufacturer of the tape. Upon evaluation of the qualification information available, it was determined that there was insufficient data available to establish a quallfled life of the tape. Resolution: Write a procedure for periodic inspection of hardcast tape. This may be part of Surveillance Procedures SP 211.01B and SP 211.01E. Qualify the Hardcast Tape. Closure Requirement: RT.HVS.011 Qualification Memo Completion Date: Prior to beginning post modification testing of the CR/TSC Essential Air System (about February 15, 1987). 4C.48.30 Problem 30 Freon leak in Essential Air Handler Tracking Number: 26.0151 4C.48-15
Description:
An informal investigation of the effects of Freon 22 on habitability has indicated that the Freon in its gaseous form would act as an asphyxiant by displacing oxygen in the Control Room, however, a very high concentration of Freon is regaired to endanger habitability. In af.,cion, discussions with other plants having a sim iar type of HVAC system has revealed that were analyses were performed, it was found no subsequent actions (i.e., installation of Freon Detection Systems, etc.) were required. Resolution: Perform a study to document the affect of a Freon leak on Control Room habitability. Closure Requirement: Study Completion Date: Prior to initial plant heat up. 4C.48.31 Problem 31 Cutrol Room /TSC Essential HVAC System (Diesel Exhaust into HVAC Intake) Tracking Number: 19.0010.G
Description:
During certain weather conditions, the exhaust of the existing diesel generators is dispersed over the Auxiliary Building roof area, including the Control Room /TSC Essential HVAC Unit outside air intakes. Concurrent operation of the diesel generators and the HVAC system could result in deterioration of the carbon absorbent material in the filter banks. Resolution: Perform a study to determine what possible long term effects the diesel exhaust may have on the carbon filters. Closure Requirement: Study affect of diesel exhaust fumes on Cr/TSC Essential Flitration Unit filters. Completion Date: Prior to initial plant heat up. l l 4C.48.32 Problem 32 The affect of CR/TSC Essential Filtration Unit initial air flow rates on CR/TSC radiation exposure levels is l not known. f Tracking Number: 26.0020 0 4C.48-16 t ~
Description:
Upon system actuation, air flow rates thru the CR/TSC p Essential Air Filtration Units exceed the stabilized flow rates for several minutes. During this period, the concentration levels of airborne radioactive particulate and or gas is assumed to be at maximum . levels. The combination of the flow rates and concentration levels may result in increased CR/TSC occupant exposure levels. Resolution: Adjust the Cr/TSC Essential Air Filtraticn Unit flow controllers to minimize the time required to stabilize air flow. Closure Requirement: HR xxxxxx Completion Date: Prior to initial plant heat up. 4C.48.33 Problem 33 Automatic initiation of Cr/TSC Essential Air System i operation increases Control Room Operator work load during certain critical plant operating phases. l Tracking Number: 26.0019
Description:
Both trains of the Cr/TSC Essential Air Conditioning System are automatically started upon Safety Features m initiation. The Control Room Operator must direct attention to this system, decide which train to shutdown, then take action. This is believed to be an unnecessary contribution to Control Room Operator workload during critical phases of plant operation. Resolution: Remove the CR/TSC Essential Air System (both trains) automatic start capability from Safety Features. Closure Requirement: ECN xxxx Completion Date: Prior to initial plant heat up. 4C.48.34 Problem 34 Temperature in the Essential Unit compressor compartments may become excessive during hot weather operation. Tracking Number: 26.0015
Description:
Ventilation in Essenb al Unit compressor compartments is natural circulation. During hot weather operation, i temperature in the compartments are elevated when the 4C.48-17
system is in operation. The elevated temperatures may result in failure of the motor, compressor, or other component. Resolution: Provide a vent between the compressor compartment and the condenser plenum. Closure Requirement: ECN-R-il70 completed Completion Date: Prior to beginntrg po'.t modification testing of the CR/TSC Essential Air S): tem (about February 15, 1987). 4C.48.35 Problen 35 Quality of the existing CR/TSC Essential Air System refrigerant compressor motors is questionable. Tracking Number: 19.00ll.H
Description:
The compressor motors are CDP motors. Concern has been expressed about the quality of the bearings, shaft keyway, and motor windings. Consideration should be given to upgrading these motors to TEFC or Mill and Chemical motors. Resolution: Provide specifications or written report about compressor motors. Closure Requirement: Specification or written report Completion Date: Prior to initial plant heat up. 4C.48.36 Problem 36 Improper system start during " dead" bus transfer. Tracking Number: 26.0483
Description:
The CR/TSC Essential Air System Train B is powered from 480V Nuclear Service Bus 382. The 382 bus is powered from the 4160V Nuclear Service Bus 482 via breaker 52-38202. The 382 bus may also be powered by the 480V Nuclear Service Bus 3B thru breakers 52-3821 and S3B2 (tie breaker). During a " dead bus" transfer, breaker 52-38202 is opened, 52-3821 is closed, and S3B2 is closed. During the evolution, the CR/TSC Essential Air Train B automatically started when S382 was closed. The train also automatically started during a reverse of the aforementioned sequence when breaker 52-2B202 was closed. O 4C.48-18 I
Resolution: Test CR/TSC Essential Air Dead Bus Transfer and repair discovered circuit flows. Closure Requirement: Test results Completion Date: Prior to initial plant heat up 4C.48.37 Problem 37 Condensate drain lines on Essential Air Handler (AH-A-545A and AH-A-5458) are potential bypass paths around the HEPA and carbon filter banks. Tracking Number: 26.0484
Description:
The Essential Air Handler condensate drain lines, which provide a drainage path for condensation from the DX cooling coil, are terminated on the Auxiliary Building roof. Although the drain lines contain liquid traps l and check valves, there is no assurance that leakage I into the essential air handler will not occur through the path. (A similar problem was identified at the Trojan Nuclear Power Plant.) Resolution: Modify the condensate drain lines Closure Requirement: ECN-R-1260 completed Completion Date: Prior to beginning post modification testing of the CR/TSC Essential Air System (about February 15, 1987). I 4C.48.38 l Problem 38: Refrigerant lines inside of CR/TSC Essential condensing units for U-545 A/B are not insulated. Tracking Number: 26.0425
Description:
The refrigerant lines and oil separator are excessively hot and restrict maintenance in the cabinet. The oli separator also acts as a heat exchanger which can cause refrigerant slugging into the compressor. Resolution: Insulate hot gas bypass, suction return line, condensor and compressor discharge lines and the oil separator vessel. Closure Requirement: HR Completion Date: Prior start up. 4C.48-19
4C.49 Component Cooling Water (CCW) 4C.49.1 Problem 1 Casualty Procedure C.20 does not provide guidance for controlling the plant after a total or partial loss of CCW. Tracking Number: 26.0175
Description:
In the event of a total loss of CCW, Casualty Procedure C.20 only instructs the operators to trip components cooled by CCW, then refers them to OP B.4 for a natural circulation cooldown. Loss of a single CCW pump would require removal of loads not necessary for plant control and might necessitate a power reduction, since 6th CCW pumps must operate to adequately remove heat loads. Resolution: Revise Casualty Procedure C.20, to address the order in which components are tripped. Include cooldown and seal injection rates that will maintain a constant pressurizer level during a total loss of CCW. Closure Requirements Revised C.20 procedure. Completion Date Prior to hot functional testing. 4C.49.2 Problem 2 incorrect flange gaskets have been used on the CCW piping to the RCP coolers. Many of the bolts are loose. Tracking Number: 21.0081.A
Description:
All of the CCN piping to the RCPs must conform to the design specifications. The current status of the piping joints, in many cases, does not reflect the design requirements. Resolution: Replace gaskets and torque bolts ac specified by the manufacturer or accepted standards. Closure Requirement: Completion of Work Requests 121631 - 121634. Completion Date: Prior to decay heat outage. O 4C.49-1 i
40.49.3 Problem 3 The CCW flow to the RCPs needs to be rebalanced to protect the new seals. Tracking Number: 26.0170
Description:
The CCW system has not been flow balanced since June 1974. Since then, much of the seal water piping and RCP seals have been modified. Resolution: Rebalance CCW system to meet new flow requirements. ECN A-5339 and NCR S-4857 must be completed prior to flow balancing. Closure Requirement: Completion of ECN A-5339 and NCR S-4857. Completion of CCH flow balance test procedure. Completion Date: Prior to plant heat up. O l O 4C.49-2
4C.50 Main Control Room (MCR) O 4C .50.1 Problem 1: Changes to the Main Control Room invalidate the CRDR Tracking Number: 26.0527
Description:
Many changes are being made to the Control Room prior to restart. These modificattom change the data in the CRDR Summary Report which was ser t to the NRC in December of 1985. Each modification to the MCR must be reviewed for human factors implications in accordance with the Control Room Design Review methods or the CRDR Summary Report will be invalidated. The modifications to the MCR include:
- 1. The annunciator System (15.0004.A, 15,0281, 26.0515, 26.0400, 20.0347)
- 2. 125VDC System (26.0389)
- 3. 6900VAC System (20.0546, 22.0446.A)
- 4. 480VAC System (26.0607)
- 5. Auxiliary Feedwater System (21.0004.A, 22.0003.2, 26.0185, 26.0187)
- 6. Electrical Distribution System (20.0518, 26.0359)
- 7. Emergency Diesel Generator System (20.6105)
- 8. EFIC (15.0041, 20.0028, 20.0178, 21.0004.B 21.0122)
- 9. Fire Protection System (15.0284, 21.0145, 26.0537)
- 10. HVAC System (19.0010.A, 19.0010.E, 21.0070)
- 11. Instrument Air System (26.0341, 22.0780) 1
- 12. Integrated Control System (15.0004.A, 15.0098, 21.0103, 22.0003, 23.0001)
- 13. Main Feedwater System (22.0007, 26.0357)
- 14. Multi-Plant System (24.0009)
O 15. Main Steam System (16.0001.8, 20.0027, 22.0698, 26.0327, 26.0332, 26.0337) 4C.50-1
- 16. Non-Nuclear Instrumentation System (20.0347, 22.0005, 22.0006, 26.0395, 26.0459)
- 17. Plant Computer System (15.0364, 21.0026, 26.0854, 26.0855, 26.0529)
- 18. Reactor Coolant System (26.0400)
- 19. Safety Features System (26.0648)
- 20. Vital Bus System (26.0267)
Resolution: Develop a Human Factors Program and apply the program to all MCR modifications. Closure Requirement: Program in place. Completion Date: Prior to restart. 4C.50.2 Problem 2 Priority #1 Procedure Changes. Tracking Number: 26.0710
Description:
During the QCI 12 investigations it was determined that several Emergency Operating Procedures provide the operators with insufficient information. 20.0194, 22.0446A, 15.0281, 26.0222, and 26.0391 require changes to operating procedures, but do not cause a physical change to the Control Room. Resolution: Provide clarification to the Emergency Operating l Procedures. l Closure Requirement: Revised Emergency Operating Procedures. Completion Date: Prior to restart. O 4C.50-2 l I _
.4C.51 Heaters, Drains and Vents System (HOV) 4C.51.1 Problem 1 Open Priority 1 work requests.
Tracking Number: 26.0935
Description:
- 1. HR 107674: 30226 GBI: Possible pipe leakage.
- 2. HR 111130: FT-32144: Feedwater heater drain alarm inoperative.
- 3. HR 107399: LG 31006: Root valve weld is leaking.
- 4. HR 106242: P-317A: Rear motor bearing running hot.
- 5. HR 105821: PSV-30232: Valve leaks through.
- 6. HR 105820: PSV-30236: Valve leaks through.
- 7. HR 105819: PSV-30235: "A" MSR relief leaks through. l
- 8. HR 105818: PSV-30231: "C" MSR relief valve leaks through.
- 9. HR 105470: PSV-30233: "C" MSR relief valve leaks through.
- 10. HR 102957: HDV-098: Excessive valve leakage causing loss of hot well level.
- 11. HR 102562: P-312A: Perform hot alignment check, motor overheating.
- 12. HR 80336: HDV-009: Repair Limitorque operator.
O 13. HR 116697: PSV-32225: Calibration.
- 14. HR 108281: HDV-2 (PSV-32428): Repair relief
- 15. HR 115385: HDV-235: Rework operator.
valve.
- 16. HR 114580: LV-32204: Packing leak.
- 17. HR 112570: 32123-12"-GB: Hanger Repair.
Resolution: Complete work requests. Closure Requirement: Close above work requests. Completion Date: Reactor start up. 4C.51-1
4C.52 Main Turbine and Extraction Steam System (HPT)
)
4C.52.1
~
Problem 1 Open work requests priority 1 (HPT, LPT, ESS). l Tracking Number: 26.0764
Description:
The following priority I work requests are open:
- 1. HPT: HR 108651: K-301: Throttle valve seals leak extensively.
- 2. HPT: HR 110993: K-301: Support leak repair by removing insulation.
- 3. HPT: HR 109225: HPT-625: Valve seat and stem leakage.
- 4. HPT: HR 123171: HPT-3: Investigate bent snubber on hot exhaust pipe.
- 5. ESS: HR 119514: ESS-036: Stroke valve after -
operator rework and adjust position indicator.
- 6. ESS: HR 119512: ESS-035: Stroke valve after operator rework and adjust position indicator. l
- 7. ESS: HR 119485: PSV-30228: Remove relief valve l and bench test.
- 8. ESS: HR 123273: PI-30349: Gage glass broken and n the needle is bent.
- 9. ESS: HR 115308: ESS-035: Set Limitorque 1imit (d stops.
- 10. ESS: HR 92238: J596: Extraction steam temperature computer point inoperable.
- 11. ESS: HR 82758: TE-30322: Input computer point T-586 is inoperable.
- 12. LPT: HR 119857: PDISL-30387: Barton differential pressure TG anti-motoring device requires calibration.
- 13. LPT: HR 115830: PT-30339: Computer point P-937 for hot reheat pressure is bad.
- 14. LPT: HR 114179: UV-0352: Investigation of EHC position indication logic for 'C' MSR Intercept Valve.
- 15. LPT: HR 114178: UV-30298: Investigation of 'D' MSR Stop Valve position indication.
- 16. LPT: HR 114177: UV-30296: Investigation of 'C' MSR Stop Valve positioning indication.
- 17. LPT: HR 114176: UV-30297: Investigation of 'B' MSR Stop Valve position indication.
- 18. LPT: HR 114175: UV-30295: Investigation of 'A'
' MSR Stop valve position indication.
- 19. LPT: HR 106243: UV-30295: EHC oil leak at 'A' MSR reheater stop valve.
- 20. LPT: HR 074113: K-303 A & B: Rework grease fittings on sliders.
4C.52-1
- 21. HPT: WR 96812: TJR-03: Bearing temperature recorder is always in alarm state.
Resolution: Complete the work requests. Closure Requirement: Close above work requests. Completion Date: Reactor start up. O l O 4C.52-2
4C.53 Main Turbine Electro-Hydraulic Control Systems (EH0) O 4C.53.1 Problem 1 Priority 1 work requests. Tracking Number: 26.0785
Description:
The following work requests were open as of 10/20/86:
- 1. HR 119042: Change out earth filter.
- 2. HR 119043: Change out cellulose filter.
- 3. HR 106368: Check EHC system accumulator pressures.
- 4. HR 12048* T-862: Sample EHC fluid chemistry and drain and inspect EHC reservoir.
Resolution: Complete work requests. Closure Requirement: Work request complete. Completion Date: Prior to RX critical. O I 4C.53-1
4C.54 Secondary Chemical Addition System (SCA) O 4C .54.1 w Problem 1 Priority 1 work request Tracking Number: 26.0771 -
Description:
There are 13 work requests reviewed (11-5-86) for items - that must be completed. Resolution: Perform the work as required to complete the work requests. - I Closure Requirement: Five hrs are for valves HR# Description , 119027 Repair UV-74004, the Chlorine Pressure Regulator l 123013 SCA-492 has a packing leak 123055 PSV-33612 has a thread leak l 108911 SCA-087 will not open or close t 109720 FV-75602D (CR-11) has a stem packing leak. 1 One HR for a pump. HR# Description l 114621 P-744B will not operate. Four hrs are for instrumentation: HR# Description 109736 FI-75803 does not indicate flow 112757 FI-75805 sticks, does not read flow properly ' 113744 AE-73907A/B remove circ. water chlorine analyzer 116013 LSH-75502 needs to be replaced Three are miscellaneous: l HR# Description 116784 Supports and Protect the Chlorine Cylinder's Pig Tail Line 116694 Build a safety shower at T-762 118515 75517-1/2 Hill Line is plugged between FV-75602A and FI-75601 Completion Date: Prior to steaming secondary plant. O 4C.54-1
4C.54.2 Problem 2 Improperly sized chlorine add'n sys. Tracking Number: 26.0884 .
== Description:== The present chlorine addition system is in need of maintenance that will cost more than replacing the system with a new more efficient one. Resolution: Replace the chlorine addition system with a properly sized unit. Closure Requirement: ECN closed. Completion Date: Prior to steaming secondary plant. O O 4C.54-2
4C.55 Plant Cooling Water System (PCW) 4C.55.1 j Problem 1 Review of open work requests Tracking Number: 26.0674
Description:
There is one work request that is open on the PCW system. Resolution: Repair this iten as required by the work request. Closure Requirement: Complete WR# 108818 air temperature indicator / alarm which give motor temperature on PCW Pump "A" is not functioning properly. Completion Date: Prior to steaming secondary plant. O O 4C.55-1
4C .56 Nitrogen Gas System (NGS) O 4C.56.1 Problem 1 Nitrogen Gas System (NGS) Work Requests closed by start up. Tracking Number: 26.0644
Description:
This Work Request applies to SFV-92520 which supplies nitrogen to the primary Pressurizer Relief Tank (V-219). WR 1157113 Stroke and time per SP.205.07A. Note: Duplicate of HR 117984 Resolution: Complete work and close Work Request. Closure Requirement: Complete WR 115713 and WR 117984 Completion Date: Prior to Hot functional tests. 4C.56.2 Problem 2 Priority 1 Work Requests for NGS. Tracking Number: 26.0801 O
Description:
Open Priority 1.brk Requests on NGS. 123023 PI-92535 out of calibration, repair as required. 121449 PI-65001 out of calibration. 115130 (PCV-92516) Clogged regulator. 118633 (PCV-92526) Adjust setpoint from 7 psig to 3 psig per AP.161-3. 113577 (SFV-92520) Suspect disconnect (DSAS-3) to be corroded due to closed light cycling. 118317 (NGS-3) Nitrogen valve in cold lab leaks by. This work request is a safety concern. Resolution: Complete work requests Closure Requirement: Complete HR 123023, 121449, 115130, 118633, 113577 and 118317 Completion Date: Prior to restart (criticality). O 4C.56-1
4C.57 Hydrogen Gas System (HGS) (h 4C.57.1 Problem 1 Priority 1 work requests Tracking Number: 26.0828 i
Description:
Work Requests that are required to be completed prior l to start up, 118714 (MI-84510): Dewpoint indication is offscale high. Investigate and calibrate as required. I 107916 Install Hydrogen spool piece during start up then requested by Operations. 108204 Install new Hydrogen bottle fill connection to match truck connection. Resolution: Maintenance will complete work requests as required. Closure Requirement: -Compiete WR 118714,107916, and 108204 Completion Date: Prior to criticality. O O 4C.57-1
( 40 SYSTEM REVIEW AND TESTING PROGRAM Rancho Seco has instituted a Systems Engineer program modeled after INPO Good Practice OP-209. Individual engineers are assigned N specific systems for which they have the responsibility to know, . among other things, the design basis, system limits'and precautions, > and to monitor the system condition. In addition, they aid other engineers, technicians, trainers and operators' when' interfacing with their assigned system. For the purposes of the Action Plan, the System Engineer receives all recommendations from the RRRB which are relevant to their assigned system. The Systems Engineer considers the' recommendations in light of their effect upon the system, performs additional investigation as required to assure the recommendations are sound , and if necessary, modifies the recommendations to present an integrated system solution to the PAG. The System Engineer develops the system testing program which will demonstrate that systems perform those functions important to the safe operation of the station. Justification will be provided for those safety significant system functions not tested prior to Restart if further. testing is not warranted. Testing which is developed by this program is expected to-demonstrate the material readiness of any system whose functional capability may be questioned. The testing will also provide an opportunity to refamiliarize the operating staff with system and plant operating conditions and procedures. The District has sent qualtfled representatives to the Davis-Besse and Three Mile Island facilitates to review their restart testing programs. Lessons learned from these visits are being factored into the Rancho Seco System Review and Test Program (SRTP). . In addition, the results of the Davis-Besse.SRTP are being reviewed for applicability to the Rancho Seco SRTP by the various system engineers. A comparison of the major features of the Rancho Seco and Davis-Besse programs is included as section 40.7. 40.1 Purpose The Action Plan incorporates a System Review and Testing Program (SRTP) whose objective is to demonstrate that systems important to safe plant operation are ready to perform their required function when Rancho Seco is ready to return to power. The specific objectives of the SRTP are as follows: Evaluate all system problems identified by the Plant Performance and Management Improvement Program (PP & MIP). Develop an integrated program of corrective actions for implementation both prior to and after Restart which will address system problems. O 40-1
s t s ,
.f Implement those corrective actions required prior to Restart.
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t-Identify those systems which require spcial conside, ration under the SRTP.
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Identify system functicas important to the safe operation of Lg 9, the plant. Develop and implement a testing program which will demonstrate those functions important to safe plar.t operation. 4D.2 Organization The SRTP Organization estaolished to perform the review of system ' problems and the development of the testing program is shown in figure 40-1. With the test program defined,'the organization will shift to'the test implementation configutation shown in figure 40-2. k
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40.3 Responsibilities 9 40.3.a System Review and Test Program Director (G The System Review and Test Program Director is responsible to the . l Plant Manager for the development and implementation of the program to assure component, system and plant material conditions can
. support safe and reliable operation. In this capacity he works with the Nuclear Department Managers, and their designees, to ensure that the system reviews are thorough, test objectives are proper, and that the. test procedurns are developed, performed and evaluated in accordance with plant procedures. This position will work closely with'the Implementation Manager to develop and maintain a detailed schedule of work activities in support of the test program.
The primary purpose of this position is to structure a comprehensive system review and test program, including: Defining the necessary organization and resources to accomplish the objectives. Developing administrative controls for the system review and test program. As a minimum, these administrative controls will address responsibilities and authority, scope of system review, test procedure requirements such as review and approval cycle, content, format, rules for test conduct, and the evaluation and reporting of results. Implementing the system review to develop corrective actions. b Developing the system testing prograal to test system design changes and system functions important to safe plant operation. Conducting the testing program, resolving deficiencies, and reviewing the test program results. 40.3.b Test Review Group 4 The Test Review Group shall be established as a Test Subcommittee under Section VI.k of the Plant Review Committee Charter. The Test Review Group shall contain the following members: Nuclear Plant Department, Chairman l Nuclear Engineering Department Member Babcock & Wilcox Representative Member Bechtel Design Member Quality Assurance Department Member Training Department Member 40-5
The responsibilities of this Test Review Group include the following: 9l Review of the Test Identification section of the System Status Report to confirm that proposed testing will demonstrate all system functional requirements important to safe plant _ operation. Review of all Test Specifications to assure that the scope and methodology demonstrate the system functions. Review of related Special Test Procedures and new or revised Surveillance Procedures and recommend disposition to the Plant Review Committee. Review of all restart related test results. The TRG will review all items in formal session and document their activities using meeting minutes and document review forms. 40.3.c Performance Analysis Group The responsibilities of the Performance Analysis Group (PAG) to the System Review and Test Program are encompassed by the duties previously described in section 2.3.2. The PAG is responsible for the review and approval of the system related modifications, and for the testing recommendations developed by the SRTP. As described in subsequent portions of this section, the PAG will review and approve: The integrated program for resolution of system issues The system functional criteria to assure functions important to safe plant operation are appropriately tested The system restart testing program The system operational readiness determination O 4D-6
l l 40.4 System Selection for Systems Review ; l' In order to expedite the process of systems review, it was decided (Ve) to identify those systems important to safety and begin a review in advance of completing the thorough and comprehensive investigative . process of the Plant Performance and Management Improvement Program i (PP&MIP) process. All systems at Rancho Seco are being investigated to some degree. These have been divided into two categories: Selected Systems - which comprises 33 identified systems, and Additional Systems - which includes the 45 remaining systems. The results of the investigations will be documented in a System Status Report (SSR) for Selected Systems, and in a System Investigation Report (SIR) for Additional Systems. ' The criteria utilized to identify the Selected Systems by the SRTP Director, which were then recommended to the PAG for treatment as
" Selected Systems" as follows:
1 A history of significant or recurring problems.
- 2. The system was related, or contributed, to the 12/26/85 event.
- 3. The system is being significantly modified.
- 4. The system has significant potential for initiating or adversely affecting transients.
d The criteria for selecting the Additional Systems are as follows:
- 1. The QCI-12 input phase has produced a recommendation for the system.
- 2. An open Hork Request existed against the system as of 07/01/86.
The program allows for upgrading Additional Systems to Selected Systems. Based on the quantity and significance of issues raised by the PP&MIP process, Additional systems may be upgraded by the PAG to Selected Systems status. The selection mechanism for determining which systems require a systems review is the review process of the Plant Performance and Management Improvement Program (PP&MIP) discussed in Section 4A. This process identifies component and system problems based upon extensive review of Rancho Seco performance history, condition, and design as well as industry experience. The system related output of the PP&MIP are fed through the RRRB to the appropriate System Engineers. On the basis of this evidence the System Engineer may recommend Additional Systems for upgrading to Selected Systems Status. O 40-7
- . ~ _.
The criteria for upgrading systems is that significant problems have been identified or a large quantity of comparatively minor problems have been identified. This upgrading is initiated by the SRTP Director or by the System Engineer making a recommendation in the System Investigation Report (SIR). This is submitted to the Performance Analysis Group (PAG) who makes the final decision on upgrading. The Systems Review and Test Program Director has proposed, and the PAG has approved, the following lists of selected and additional systems. The current listing of selected systems is as follows:
- 1. Reactor Plant Reactor Coolant System Decay Heat Removal System Seal Injection and Hakeup System (including High Pressure Injection)
Purification and Letdown System Nuclear Service Cooling Water System Nuclear Service Raw Water System
- 2. Steam Plant Steam Generator System Main Steam System Main Feedwater System Auxiliary Feedwater System
- 3. Power / Control Reactor Protection System (including ARTS)
Safety Features Actuation System EFIC (originally Main Steam Line Failure Logic) Integrated Control System Non-Nuclear Instrumentation System 12S Volt DC Vital Power System 12S Volt DC Non-Vital Power System 120 Volt AC Vital Power System 480 Volt AC System 4160 Volt AC System 6900 Volt AC System Radiation Monitoring System *
- 4. Auxiliary Control Room /TSC Essential HVAC System Emergency Diesel Generator System Component Cooling Water Plant Air System Instrument Air System Auxiliary Steam System
- Fire Protection System
- NSEB Essential HVAC*
Reactor Sampling System
- Reactor Building Atmospheric Systems
*Upgradad System 40-8
The current listing of Additional Systems is as follows:
- 1. Reactor Plant V.
- Borated Water System Core Flood System Containment Building Spray System Control Rod Drive System Pressurizer Relief Tank System Spent Fuel Cooling System Reactor Coolant Drain System
- 2. Steam Plant Air Ejector System Extraction Steam System Gland Steam Condenser System Heater Drains and Vents System l High Pressure Turbine System Low Pressure Turbine System Main Condensate and Makeup System
- 3. Power and Control Annunciator System I
Cathodic Protection System Plant Communications System Electro-hydraulic Control System Main Generator System Nuclear Instrumentation System p) ( - Plant Computer System Plant Security System Seismic Monitoring System Turbine Supervisory Instruments
- 4. Auxiliary Drainage and Sewage System Cranes and Holsts Carbon Dioxide System Diesel Fuel Oil System Demineralized Water System Fuel Handling System Hydrogen Gas System HVAC - Other Lube Oil System Main Circulating Water System Nitrogen Gas System Plant Buildings and Structures Plant Cooling Water System Radwaste System Secondary Chemical Addition System Seal Oil System Site Reservoir System Service Water System Turbine Plant Cooling Water System Turbine Plant Sampling System O
- Waste Gas System 40-9
40.5 System Review Report Summary The review of the Selected and Additional systems under the System Review and Test Program will be documented in individual system reports. A System Status Report (SSR) will be prepared for each Selected System aad a System Investigation Report will document each Additional System Review. The process for the development, review and approval of these reports is diagrammed in Figure 4D-3. 40.5.a System Status Report Overview The SSR for Selected Systems will be developed in three stages, with each stage documented in a separate revision (Rev. O, 1, and 2). The purpose of the Rev. O report is to initiate plant design and modification work. The contents of the Rev. O report is an Executive Summary, a basic System Functional Description, and a listing of Problem Statements developed from the PP&MIP process and from the results of a review of open Work requests. This Rev. O report is reviewed and approved by the Performance Analysis Group and the Deputy General Manager, Nuclear. Once it is approved, it is used by the system engineer to initiate design activities, maintenance activities, and plant modifications through conduct of a
" kickoff meeting".
The Rev. 1 SSR is utilized to identify tha testing necessary for each Selected System. It builds on the Rev. O SSR, providing a more detailed System Functional Description, additional problem statements from a review of open ECN's, open NCR's, and outstanding abnormal tags, and the identification of testing required to demonstrate functions important to safe plant operation to be conducted prior to Restart. Rev. 1 is reviewed and approved by the Test Review Group, Performance Analysis Group, and Deputy General Manager, Nuclear. Once approved, the preparation and implementation of test specifications and procedures will begin. A Rev. 2 SSR is prepared for each Selected System and is utilized for final system acceptance. This report contains everything from Rev. I plus additional problem statements documenting the results of the system walkdown, review of maintenance history trend investigation, and a review of the Davis-Besse SRTP results. This revision will also contain a summary of results of tests performed l to date and an operability statement by the System Engineer. This i report is reviewed and approved by the Test Review Group, Performance Analysis Group, and Deputy General Manager, Nuclear prior to Restart. 4D.S.b System Investigation Report Overview System Investigation Reports (SIR) will be developed for all ! Additional Systems in two stages. The Rev. O SIR for each system I will be utilized to get plant work started and to determine whether the system should be upgraded to Selected System status. The contents of the Rev. O report are an Executive Summa.y, a System 40-10 l
L Functional Description, and a listing of Problem Statements n developed from the PP&MIP process and from the results of a review (j) I of open Hork Requests, open NCR's, outstanding abnormal tags and open ECN's. The report also contains a justification for upgrading the system to Select Status or for maintaining the Additional System . status. This report.is reviewed by the PAG and by the Deputy General Manager, Nuclear. Rev. 1 of the SIR for each system is utilized for final system acceptance and for consideration for upgrading to Selected System status. The contents of the Rev. I status report contains the Rev. O SIR plus any additional Problem Statements and an Operability Finding. This report is approved by the Performance Analysis Group and the Deputy General Manager, Nuclear. 40.5.c Details of Report Components The following provides additional details of the major components of the SSR's and SIR's. 40.5.c.1 System Functional Description The System Functional description is a listing of the capabilities that the system must provide in order to assure reliable operation and effective accident mitigation. The functions determined to be important to safe plant operation will be identified for evaluation in the test identification process of the SSR. The System Engineer will prepare the System Functional Description with the support of the System Design Engineer who will approve the list of functions. Source documents to be used are: USAR Technical Specifications Emergency Operatin7 Procedures B&W Guide Specifications Bechtel Rancho Seco Design Manual NEP 5400 (Systeta Design Bases) Vendor manuals Applicable design calculations. Comparison will be made to the Davis-Besse function determination. 4D.5.c.2 Resolution of System Preblems The System Engineer will document all problems with the system via the Problem Statements. These problems are identified via the PP&MIP program (RRRB recommendations) and by several other processes described below. Resolutions are proposed for each problem. Priority is indicated as Restart, near term or long term. The System Engineer considers the combined effect of the resolution
% on the system and deterinines if all known system deficiencies are ,'j corrected. He will also satisfy himself that no new issues are created.
40-11 _ _ _ _ _ _ _ _ _ _ _ _ ~
I In addition to problems identified through the PP&MIP the following processes will also provide input to the system engineer. Problems identified through these activities will be documented Problem Statements. Review of Open Work Requests Open Nork Requests will be reviewed and those which should be completed prior to Restart will be identified. Those Work Requests which may be deferred until after Restart will also be identified on a Problem Statement and prioritized accordingly. Review of Open Engineering Change Notices (ECN) Open ECN's will be reviewed and be classified as requiring implementation either pre or post restart. Review of Open NCR's Open NCR's will be evaluated to identify those requiring disposition prior to Restart. Review of Outstanding Abnormal Tags Abnormal Tags, used for temporary plant modifications, will be reviewed to identify those that should become permanent modifications and those which should be cleared prior to Restart. Review of Maintenance History Trends The System Engineer will review trends of maintenance history for indications of system deficiencies. Review of Findings from Davis-Besse The System Engineer will review the system / component problems identified at Davis-Besse as documented in the Davis-Besse System Review and Test program for applicability at Rancho Seco. Material Condition Walkdown The Systems Engineer, System Design Engineer and a designee of the Operations Manager, will inspect the system for evidence of a deterioration in material condition. A checklist will be provided to assure consistency in the conduct of these , walkdowns. Problems found during the walkdown will be reported l in the System Status Report and resolved using normal l corrective action processes such as Work Requests and ) Nonconformance Reports. 40.5.c.3 Test Identification l The System Engineer will review the testing that is being performed on the system. A comparison will be made to determine if all of the functions important to safe plant operation defined in the System Functional Description are being demonstrated by current testing. If the functional testing is found to be inadequate or incomplete, appropriate additional testing will be specified. 40-12
f- The results of the test identification will be documented in the l- System Status Report Rev. I and will specify the scope of Restart testing to be conducted. Tests used to demonstrate system function will be included or referenced.
~
Justification will be provided for those functions or portions of functions not to be tested prior to Restart. 40.5.c.4 Operability Finding i Based on the identified testing program and the recommendations and dispositions for system problems the Engineer will specify any additional maintenance, modifications, procedure changes, and testing that is needed to assure reliable system operation. He will indicate how these open items are being tracked to completion. O . O 4D-13
4D.6 Restart Test Program Testing to be performed prior to Restart will be identified by the SRTP and by other normal testing programs. The scope of testing considered part of the Restart Test Program includes the functional i testing identified by the SRTP, post modification testing not f covered by the SRTP functional testing, and the series of tests already identified below:
.1 Loss of ICS/NNI Test
.2 Loss of Instrument Air Test - Both a sudden and a gradual loss of instrument air (similar to recommendation in Reg. Guide 1.68.3) will be simulated with the plant on steam generator cooling (subject to the resolution of any unreviewed safety questions). 1
.3 Integrated Leak Rate Test of Reactor Building.
.4 Complete the Balance of Ten-year In-service Inspection (except for those inspections requiring removal of the reactor vessel head). This includes required periodic pump and valve testing. ,
.5 Integrated Engineered Safeguards Actuation Test
.6 Emergency Diesel Generator Biennial Inspections The Restart Test Program will be implemented by the SRTP organization identified in section 40.2 and shown in figure 40-2.
The responsibilities of the SRTP Director, the TRG, and PAG with , respect to the Restart Test Program have been delineated in section 40.3. The responsibility for the preparation and conduct of new test procedures belongs to the System Engineers. A representative of the SRTP organization will be assigned as a test leader for the conduct of new procedures supporting the Restart Test Program. The Restart Test Program is controlled by a series of formal procedures. Most activities are controlled by the normal plant procedures while new procedures have been created for the overall control of the program and other activities peculiar to the SRTP. The procedural controls considered as part of the Restart Test Program address the following activities: Development of the functional testing program scope. Preparation, review and approval of test l abstracts / specifications. l Preparation, review and approval of test procedures. Conduct of Tests. Review and approval of test results. Training and certification of test personnel. l Safety Tagging. Control of test equipment and instrumentation. Control of temporary plant modification. 40-14
l
- Identification, control, and reportability of non-conforming i conditions.
Quality Assurance / Quality Control. !~ The results of the Restart Test Program will be reviewed by the System Engineer and by the Test Review Group. The results of those tests completed prior to the development of Revision 2 to the System Status Report will be described in the report to be considered in the evaluation of system suitability for return to service, f
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O . O 40-15 l
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40-16 0
-40.7 Comparison Between Rancho Seco and Davis-Besse Programs This chart is a comparison between the Rancho Seco and Davis-Besse CT System Review and Test (SR&T) program. The significant difference between these programs is that Rancho Seco will do a partial system review for all remaining systems and consider them for possible upgrading to Selected Systems if they meet the criteria as established.
1 lO l O 40-17
SYSTEM REVIEW AND TEST PROGRAM RANCHO SEC0/ DAVIS-BESSE COMPARISON
.1 System Selection Rancho Seco Davis-Besse Selection criteria includes
- Selection criteria includes non-safety systems non-safety systems 27 systems initially selected
- 31 systems initially selected Process for inclusion of ad-
- During course of program ditional systems established - 3 additional systems in-
- 6 systems added to date cluded in program Partial system reviews to be
- Systems outside program performed on all remaining deferred for subsequent systems with any identified review, problem or outstanding Maintenance Work Orders.
O O 40-18
J This chart provides a comparison between the Rancho Seco/0 avis-Besse problem
- identification process.
b' SYSTEM REVIEW AND TEST PROGRAM RANCHO SECO/ DAVIS-BESSE COMPARISON
.2 Problem Identification Rancho Seco Davis-Besse System based' problem identification
- System based problem identification
- Review of outstanding work orders - Review of outstanding work orders
- Review of open ECN's - Review of open FCR's (ECN's)
- Review of maintenance history trends - Review of maintenance history trends
- System walkdown for material - System walkdown for material condition condition
- Input from day-to-day system - Input from day-to-day system responsibility responsibility
- Review of test results of comparable Davis-Besse system
- Review of Licensee Even Reports
- Review of significant DCRDR - Review of significant DCRDR Human Engineering Discrepancy Human Engineering Discrepancy Reports Reports
- Review of Transient Assessment Reports
- Input from Operations and j Maintenance personnel Programmatic based problem identification includes:
- Precursor review
- Plant staff interview
- Deterministic failure consequences
- B&WOG Safety and Performance Improvement Program
- Other PP & MIP activities i
I l l l
\
40-19
The following chart provides a comparison between the Rancho Seco/ Davis-Bet-system review process. SYSTEM REVIEW AND TEST PROGRAM RANCHO SEC0/ DAVIS-BESSE COMPARISON
.3 System Review Rancho Seco Davis-Besse System level functions to be
- System level functions to be identified for each system identified for each system Problem identification both Problem identification system and programmatically system based based Corrective actions developed
- Problem prioritization criteria
! and recommended by System by System Engineers Engineers and PP & MIP Problem report review and
- Problem report review and corrective action approval corrective action concurrence by PAG by IPRC O
O 40-20
This chart provides a comparison between the Rancho Seco/ Davis-Besse test program development.
/s (s_,/ SYSTEM REVIEW AND TEST PROGRAM RANCHO SECO/ DAVIS-BESSE COMPARISON
.4 Test Program Development Rancho Seco Davis-Besse Identify new test requirements
- Identify new test requirements for functions not adequately for functions not adequately tested tested Identify retest requirements
- Identify retest requirements for functions previously tested for functions previously tested or provide justification for or provide justification for why retest not required why retest not required
- Identify new test requirements
- Identify r.ew test requirements
; to support plant modifications to support plant modifications Review and approval of test
- Review and approval of test requirements by TRG and PAG requirements by IPRC O
3 i I 40-21
This chart provides a comparison between the Rancho Seco/ Davis-Besse restart testing program. SYSTEM REVIEW AND TEST PROGRAM RANCHO SEC0/ DAVIS-BESSE COMPARISON ,
.5 Restart Testing Rancho Seco Davis-Besse Test specifications and test
- Test specifications and test procedures to be prepared procedures to be prepared under existing procedural under existing procedural controls controls Test specifications and test
- Test specifications and test procedures to be reviewed procedures to be reviewed by TRG by JTG Test leaders to be trained
- Test leaders to be trained and qualified on station and qualified on station testing program requirements testing program requirements All test results required to All test results required to verify function to be reviewed verify function to be reviewed by TRG by JTG & IPRC O
O 40-22
APPENDIX A DISTRICT BOARD OF DIRECTORS' POLICY STATEMENT ON PERFORMANCE IMPROVEMENT AT RANCHO SECO
Rev. 8 APPENDIX A DISTRICT BOARD OF DIRECTORS' POLICY STATEMENT ON PERFORMANCE IMPROVEMENT AT RANCHO SECO I. INTRODUCTION On Jul) 3,1986, the Board of Directors of Sacramento Municipal Utility District voted unanimously to adopt the policy and the associated performance improvement goals presented below. This policy and related planning guidance are intended to provide direction to all persons who may become involved in this effort. The policy and associated implementation guidance stated herein are appropriate in view of the increased internal and external emphasis being placed on plant performance improvement. The emphasis on performance improvement his special significance for Rancho Seco for two reasons; it is a B&W plant which is perceived by the NRC to be more sensitive to upset conditions than other PHR's, and among the B&W plants, it has the poorest overall performance record. To put the Performance Improvement Policy Statement below in proper perspective, it is appropriate to restate the District's primary mission. That primary mission is to generate electricity safely, reliably, economically. Carrying out this mission in a responsible manner involves a continuous emphasis on safety. It includes strict 9 adherence to the SMUD Quality Program. It also includes taking the necessary short-term actions to achieve and maintain a desired balance over the long term. In carrying out this primary mission, the District is accountable to many parties. They are accountable to: their customers-owners, their bond holders, the general public, their employees, the Nuclear Regulatory Commission, and other nuclear and non-nuclear regulatory agencies. The management challenge in establishing an optimum balance and meeting each of these responsibilities in an appropriate manner is significant. This Policy Statement is intended to aid in carrying out the actions required to significantly improve the performance of the Rancho Seco Nuclear Generating Station in a manner which is consistent with the District's primary mission. II. POLICY STATEMENT The District's Board of Directors is committed to achieving a prompt and significant improvement in performance at Rancho Seco and to provide the support necessary to achieve standards of excellence in all aspects of nuclear activities. To reinforce that commitment, the Board has G established the following objectives related to performance improvement at Rancho Seco. A-1
I 1 Near-Term District Objectives Accomplish those actions which will substantially reduce the likelihood of another significant transient at Rancho Seco. ' Initiate longer-term actions which will contribute to a sustained improvement in plant performance at Rancho Seco. Long-Term District Objectives
- Accomplish those actions which will allow the District to achieve the 1990 Performance Improvement Goals for Rancho Seco.
Performance Improvement Program Goals The 1990 goals which have been committed to INPO include the following items: Improved Equipment / Plant Availability Reduced Forced Outage Rate Reduced Reactor Scram Rate Reduced Unplanned Safety System Actuations Reduced Safety System Unavailability i Improved Thermal Performance Reduced Low Level Waste Volume Reduced Personnel Exposures Reduced Industrial Accident Rate III. POLICY IMPLEMENTATION AND PLANNING GUIDANCE The Board considers the Performance Improvement Program at Rancho Seco to be the District's highest priority activity. The Board intends to closely monitor progress toward the Performance Improvement Program goals. The near-term objectives shall be satisfied prior to the restart of Rancho Seco in accordance with the approved Performance Improvement Program. The long-term objectives shall be pursued in accordance with the approved Performance Improvement Plan. District management shall develop, maintain, and follow a detailed Program Plan which encompasses all projects related to the Performance Improvement Program. The Board of Directors encourages 01 strict participation in the activities of industry groups where sharing of information and costs can be beneficial to the District. This is especially appropriate regarding participation in the B&W Owners Group activities which involve interactions with other utilities with plants of similar design. O A-2
A, -m_ _ k 4 APPENDIX B SPECIFIC OISTRICT RESPONSES TO ' NUREG-1195 FINDINGS O 4 J 1 O i'
APPENDIX B b SPECIFIC DISTRICT RESPONSES TO ; h NUREG-1195 FINDINGS The following are the District's responses to the findings and conclusions section of NUREG-il95. FINDING - SECTION B.1: ICS/NNI Power Supply Monitor
- 1. The December 26, 1985 overcooling transient was initiated by the power supply monitor in the nonsafety-related ICS (tripping the
+/- 24 Vdc power). The most probable cause of the tripping was a design weakness that apparently made the circuit susceptible to erratic operation if " contact resistance" between the 24 Vdc bus and the power supply monitor were to develop, and the
~ development of a high resistance connection (i.e., a bad crimp l connection) in the wiring between the + 24 Vdc bus and the power
~
supply monitor which exposed the design weakness and caused the module to trip. (SMUD has agreed to further explore the cause d the failure of the power supply monitor by having an independent laboratory conduct additional analyses). 4 I DISTRICT RESPONSE Evidence obtained after the December 26, 1985, trip supports the District's belief that the power supply monitor did not fail, but, in Q y fact, performed its required function. Test results by Scientific Application International Corporation (SAIC) have shown that, although the PSMs that was installed in the ICS did have a cold solder joint, this did not cause the 12/26/85 trip. A recommendation that redundant PSMS be incorporated into the design was rejected based upon the excellent service record of the units (no failures sto operate, or inadvertent operations, in approximately 100 service years). Further, SAIC tests were performed on the Sl/S2 switches from the ICS. The switches were found to trip in only 0.14 to 0.16 seconds.
- The switches were to have approximately 0.5 seconds delay between the time the PSM demanded a trip and the time the switches actually opened, in order to reduce the number of trips due to short duration power losses. No design changes are deemed necessary, because operating experience has shown no spurious trips have occurred due to reduced time delays. Also during SAIC testing, one of the switches was disassembled and inspected, and found to be very clean inside, with no damage or deterioration evident.
While 1) the PSM sensitivity to external resistance in series with its voltage input, and 2) the short time delay characteristics of the
- Sl/S2 switches, have been identified as contributory causes to the 12/26/85 trip, the direct cause was identified as the loss of ICS DC power caused by a manufacturing error on a lug improperly installed l on a factory prepared wire.
B-1
Specific plant modifications were engineered and installed to address the identified lessons learned. Engineering Change Notices and subsequent field work accomplished the following:
- a. Leads to the power supply monitor (PSM) now go directly to the power supply bus on the ICS and NNI. This eliminates unnecessary voltage drops on the PSM sensing leads due to the external resistance developed by PSM previously being included in a " daisy chain" that provided power to 23 additional electronic modules.
- b. Inspection and correcti3n of terminations (i.e., lugs) has been completed. Over 40,000 terminations throughout the ICS, NNI, SFAS, and RPS cabinets supplied by Bailey Meter Company have been subsequently inspected / tested / upgraded. While no generic problem was found, this effort should significantly improve the confidence in the hardware reliability.
- c. ICS Power Supply Cabinet Bus wiring was replaced.
- d. A new power supply monitor and new Sl/S2 switches have been calibrated and installed.
New battery and diesel generator backed inverters are being provided for both the normal and standby supplies to the ICS/NNI systems. Other changes were made to the DC power supplies themselves to ensure that their current capacities have adequate margin over their load requirements. FINDING - B.2: Repositioning of ICS Contrclied Valves on Loss of ICS
- 2. Upon loss of ICS dc power and the subsequent automatic l repositioning of a number of valves in the plant, the design of l the ICS also caused the loss of remote control of the affected
! valves from the control room which necessitates manual actions locally at the valves. DISTRICT RESPONSE Design changes have been completed, or are underway, to change the operation of those valves important to mitigating the effects of loss of ICS Power. The Turbine Bypass and Atmospheric Dump valves now remain closed upon loss of ICS power, while control is passed to devices which are powered independently of the ICS. These valves can now be opened / closed by the operator from the control room. As described in Section 4C.1, the AFW flow control valves will be l automatically controlled by EFIC to maintain correct steam generator level on loss of ICS Power. Also, upon loss of ICS, EFIC will automatically maintain steam pressure via the atmospheric dump
- valves. The operator will be able to take manual control from within the Control Room. Status indicating light will be provided in the control room for the atmospheric dump valves, turbine bypass valves, and rain feedwater and startup feedwater control valves.
B-2 l
FINDING - 8.3: PM Program for Manual Valves
- 3. An AFW manual isolation valve could not be shut by the operators
[V after the failure of the auxiliary feedwater (AFW) (ICS) flow control valve. The failure of the AFW manual isolation valve was the result of a lack of any maintenance on this valve during the operational life of the plant. The lack of a maintenance program resulted in the valve being inadequately lubricated, which caused the valve to seize. It appears that the lack of a maintenance program could affect the operability of other manual valves at Rancho Seco. DISTRICT RESPONSE Troubleshooting identified a lack of lubrication and rusted yoke nut bearings on Auxiliary Feedwater Isolation Valve FWS-063. Reworking these components restored the valve to an operable condition. As an element of the troubleshooting effort, the similar valve on the OTSG-B line (FWS-064) was inspected, as were all similar valves in service on the AFW system. All were found serviceable with only normal closing torque required to operate through their full travel, , although evidence of recent lubrication was missing. These valves are maintenance isolation valves and are required to be " locked" in position during power operation. In recognition of the desirability of having certain manual valves readily operable, the Nuclear Operations Manager has identified 143 valves which will be verified operable prior to resumption of power Q(
/ operation. These manual isolation valves are characterized by their purpose which is mitigate specific casualties or both active (pumps, valves) and passive (tanks, heat exchangers, pipes) equipment. They were selected to include both primary and secondary plant systems necessary to power production or nuclear safety. Function, not class, was the selection criteria. The program involved exercising (stroking) of the valves and, where necessary, servicing with lubricants, packing, or adjustments was done and documented. The results of exercising the selected valves did not show a generic problem as the result of the previous lack of either a programmatic PH program, or periodic stroking. Valves did receive PM to ensure future operability, and one class of valves, manually operated with limitorque operators, were identified as needing overhaul. As a result, a plant-wide effort has reworked all operators in this category to assure future operability.
Significant changes are underway with respect to the Preventive Maintenance Program at Rancho Seco. Staff is being added for the specific purpose of expanding the scope, content, and quality of the preventive maintenance program. Specific procedures detailing the PMs are being provided or expanded to give confidence in the condition and operability of the PM'd equipment. This expanded PM program includes the above identified valves, as a first phase in the expansion, in addition to those already receiving periodic maintenance. O B-3 1
FINDING - SECTION 8.4: Procedure Guidance for Loss of ICS
- 4. Rancho Seco Emergency Operating Procedures (EOP) do not address the loss of ICS power. The lack of specific guidance seems to be a weakness in the plant-specific E0Ps available to the operators on December 26, 1985. The Rancho Seco Anticipated Transient Operating Guidelines (ATOG) supplied by the B&W Owners Group include an explicit procedure for a loss of ICS power and the ATOG directs operators to that procedure. However, this procedure was not included in the Rancho Seco E0Ps.
DISTRICT RESPONSE During the December 26,1985, event those actions necessary to respond to the consequences of the Loss of ICS power were appropriately defined within the symptom-based EOP's. The operating philosophy is to place the plant in a stable post-trip condition, and then begin the cause-of-event trouble shooting, e.g. ICS power restoration in this case. In this event, due to the difficulty in closing the AFW valves and reluctance to trip the AFW pumps, restoration of ICS power would have mitigated the event earlier. While a Loss of ICS Casualty Procedure would have been helpful in expediting the power restoration, that function correctly should not be a part of the symptom-based E0P's. The E0P's must address the condition when ICS Power cannot be restored, for whatever reason, and a specific Casualty Procedure has been implemented for this requirement. A review of the AT0G determined that there were no other needs er requirements which were not incorporated in the E0P's. Training has been, or will be, given to the appropriate plant staff on the lessons-learned and changes made relative to the E0P's and plant operations. l FINDING - B.5: Feedpump Trip Criteria
- 5. The E0Ps at Rancho Seco direct the operators to trip the l
appropriate feed pumps to terminate flow if the feedwater flow I cannot be isolated. This was not done during the December 26, 1985 incident. The operators were reluctant to stop the AFW l pumps even when they had difficulty stopping flow to the once-through steam generators (OTSG) by valve operation. The i operators had decided that they would stop the AFW pumps only if water started to flow into the main steam lines. However, the operators failed to adequately monitor OTSG water level and, as a result, water was introduced into the steam lines. Their reluctance appears to be the result of the substantial emphasis placed on the AFW system by NRC and others, and a lack of l confidence in the reliability of the AFW pumps (i.e., fear that l the pumps would not restart if stopped). O B-4
DISTRICT' RESPONSE !: b ~The E0Ps did not contain specific parametric criteria such as RCS
; . Temperature, Steam Generator Level, or Pressurizer Level for when to i
. trip main and auxiliary feedwater pumps during an overcooling. Lack of specific criteria let the operators.be influenced by perceptions of NRC concerns regarding auxiliary feedwater operability.
Procedures have been revised to specify when to trip main, condensate and auxiliary feedwater pumps. This is a significant improvement within the E0P's as it removes the obstruction presented to the , operator, and replaces it with a preplanned response to the observed' , conditions. 3 The E0P's and the philosophy of symptom-oriented plant control are topics addressed in both initial and continuing training for all licensed operators. Topics for re-emphasizing AFW trottling/ trip criteria were included in the Restart Training Program. Specific discussion centered on ensuring a clear understanding of the conditions under which AFW must not be throttled, throttled to prevent overcooling, and terminated. Simulator scenarios, including the interface with the EPIC system, were used to strengthen that understanding. A detailed review of the starting and operating reliability of the F Auxiliary Feedwater pumps was done which did not support a lack of. confidence in this equipment by the operators. In the twelve year 1 operating history (315 attempted starts) three instances of failure-to start these pumps were noted. In the first case, in early 1975, i
~
O P-318 tripped on over current during a Surveillance Procedure Test start. All class I motor overcurrent settings were checked as a result. The second case occurred on P-318 in 1980 when it failed to start during a surveillance test. This could not be duplicated. The l i third occurrence was a failure to start P-318 on its turbine drive, !- also for a survell_ lance test. P-319 has.always started. There has i never been a time when there has been loss of function of the
- Auxiliary Feedwater Pumps.
l FINDING - B.6: Priority of PTS or Pressurizer Level l
- 6. The operators had considerable difficulty reconciling the
! dichotomy between avoiding the pressurized thermal shock (PTS) , region [e.g, reducing high pressure injection (HPI) flow] and [ regaining pressurizer level (e.g., increasing HPI. flow in l accordance with their EOPs). Their training and procedures were I not adequate to resolve this conflict and to some extent tended [ to provide conflicting indications of the appropriate priorities. i ,1 DISTRICT RESPONSE i i The Emergency Operating Procedure (EOP) in place on December 26,
- 1985, included rules which specifically state the events or plant i conditions which mandate throttling of HPI flows. Under Section 2.2 t
it is clear that HPI flow should be throttled to prevent exceeding brittle fracture limitations. B-5
-- -- _. - . - _-_ .- __~
The training programs for all License Training have been examined. The conclusion being that HPI throttling, even with no pressurizer level, is adequately addressed through the following:
- a. Both initial License Training and Senior License Training Programs address E0P rules and provide background information.
- b. Many drill scenarios in the Simulator Training Program contain the application and use of rule 2.
- c. All E0P rules are required to be committed to memory by all control room operators and are tested during NRC License and Rancho Seco Requalification Examinations.
The " conflicting priorities" concern has been further addressed by the training given to the operators since the event, which emphasizes the purpose and requirements of the E0P rules and the hierarchy of implementation. At the same time, the E0P's, have been reviewed and enhanced to provide clear direction to the operator when faced with an apparent conflict between avoiding PTS and maintaining pressurizer level. HPI flow may be throttled anytime RCS is subcooled, regardless of pressurizer level (or lack of it). Although improvements in the ATOG derived symptom-oriented E0P's have been made as a result of this event, a comprehensive operational assessment of the E0P's demonstrated that they are an adequate and viable way of responding to the spectrum of transient events. For events such as this overcooling, they were adequate for responding to PTS concerns. The improvements have been directed toward precluding conditions which precipitate PTS issues. Training to emphasize these issues has been given to the operators. A comprehensive report describing the Restart Training Program, dated September 1986, has been prepared and is available for review. In response to questions 19(a) (1/2) in a September 5, 1986, NRC letter, the following is given:
- 1. Describe how HPI vs. PTS concerns are directly handled in the E0Ps.
DISTRICT RESPONSE l Emergency Operating Procedure Rule 2, "HPI Flow Control," has been
- revised to state explicitly that HPI should be throttled to prevent i
entering the PTS restricted region. Combined with the previously existing statement in Rule 2, that HPI can be throttled any time reactor subcooling margin is restored, there is no ambiguity regarding HPI vs. PTS concerns. Procedure steps which call for HPI initiation refer to this rule.
- 2. Why were the PTS guidelines exceeded during the December 26, 1985 event?
O B-6
4 DISTRICT RESPONSE l [d Operators observed the impending PTS violation but did not take prompt action to minimize excursions into the PTS restricted region.
.Several factors contributed to exceeding the PTS guidelines.
l a. E0P's did not provide explicit enough instructions'to assure rapid termination of the cooldown, such as an RCS minimum ,! temperature criteria for requiring feed pump trip.
- b. E0P's did not explicitly require cooldown termination from the control room, thus in-plant actions were initiated to control the cooldown. Coordination of in-plant effor ts required 4
attention of control room operators which detracted from timely recognition of impending PTS criteria violation. t
- c. E0P's did not explicitly contain direction to gain plant control ,
prior to initiating efforts to regain ICS power. Control room personnel efforts to restore ICS power detracted from timely recognition and response to the impending PTS criteria violation. 1 i
- d. The subcooling margin stopped increasing for a few minutes, i prior to rapidly increasing into the PTS restricted region, j Operator observation of subcooling during this stable period
- j. could have lead the operators to conclude the situation was i stabilizing and allowed them to momentarily direct their ;
attention to in-plant efforts or ICS power restoration. ' Revisions to the E0P's addressing the above factors have been made and appropriate training conducted. Numerous simulator exercises 1 with all crews have shown that these changes are effective in avoiding PTS. , 2. Discuss the rationale of using overly conservative PTS Guidelines which do not allow the operators to know if a real problem exists which'might require emergency pressure reduction. ; , DISTRICT RESPONSE 1 ! PTS analysis requires numerous assumptions for the many parameters i that enter into the calculations. Rather than require the operators
, to figure out the validity or applicability of each assumption,
- conservationism has been applied to each. For example, for fluid mixing (temperature of water against the vessel wall), no forced flow is assumed to exist; the neutron fluence is assumed to be that of 32
, EFPY; a flaw size equal to the worst allowed by the ASME codes is i assumed to exist; and a repressurization to 2,500 psig is assumed to aCCur. , ! The guidance given the operator, if he should enter the undesired l region, is to simply leave the region by lowering pressure. There
- are too many parameters involved to develop families of optimum l curves for each possible condition. The procedural guidance is
! conservative. I B-7 4
--- . _ _ . . . - _ _ _ - _ - - __ . _ . - , - . . - - _ _ . . . . . . . - - - _ ~
FINDING - B.7: Training on Loss / Restoration of ICS Power
- 7. The operators received neither classroom nor simulator training on the overall plant response to either the total loss of ICS dc power or the restoration of ICS dc power.
DISTRICT RESPONSE Training programs have been revised, to address the lessons learned from the December 26, 1985 event, and expanded to include the plant modifications and procedure changes which have occurred. Operator simulator training time has been increased by 607. for this year and the operators are being scheduled for two weeks of simulator training in 1987. This is twice as much as was previously scheduled. The post-event simulator training included the following items:
- a. Emergency Operating Procedures (EOP) Training including all steps necessary to terminate overcooling, or OTSG overfill from any cause, including the loss of ICS power.
- b. Effect of changes to ADV, TBV, and AFW valve operation following loss of ICS power.
- c. Command, communications, and control training, including implementation of the Emergency Plan.
- d. Recovery from SFAS, i.e., restoring normal makeup and letdown flow.
- e. Differences between the B&W Simulator and the facility (Operator traps).
- f. HPI and AFW throttling and pump trip criteria,
- g. PTS recovery actions.
- h. Cooldown rate interpretation and tracking.
- 1. Transition from AFW to MFW flow.
- j. Use of multiple indications.
- k. Operation of manual valves.
The Olstrict is in the final stages of procuring its own plant specific simulator. Not only will the installation of a simulator at Rancho Seco afford additional crew training, but the simulator will also incorporate the Control Room Design Review Human Factors modifications. The establishment of a state-of-the-art training facility will significantly enhance the quality of operator training. O B-8
i-The effects of Restoration of ICS Power are under investigation by
- ' -the B&W Owners Group. Modifications to ADV's, TBV's, and AFW Control
. Valves mean that following loss of ICS power, control will '
j automatically transfer to independent controls. Procedures cause 4
.these to be in manual mode when attempting to repower the ICS. Since ,
these controls are independent of the ICS, whatever demands the ICS ! issues will have no effect. Once stable ICS conditions are observed,
- the operator can return each device to ICS control.
i i Installation of EFIC will provide ADV and AFW control independent of l
- ICS.
j~ "0PERATOR TRAPS" - DIFFERENCES BETWEEN THE B&W SIMULATOR AND RANCHO j SECO ! 1. MUT outlet valve on 'B' SFAS panel at the simulator has a ! control on each, wtth the 'A' panel normally powered up due to l the power source selection for this valve. 1 j 2. HPI/MUP recirc valves have no controls on the SFAS panels at the j simulator. These valves are modeled to auto-close on an SFAS ! Initiation. To re-open, the simulator instructor must do them i at the simulator control panel. ;
- 3. The simulator doesn't model separate controls for.the HPI/MUP lube oil aux 111arles on HlRC. Operators must understand that procedures require these aux 111arles to first be started when starting an HPI/MUP.
j 4. The simulator allows the MVP to draw essentially all its suction from the BWST when SFV-25003 and the MUT outlet are both open. At the plant, the MUT pressure and level determine whether the MUP will draw water from the MUT, BNST, or both. There is a i check valve on the MUT outlet at the plant, but not at the l simulator.
- 5. Response of RCS (pressure and level) to changes in secondary
- pressure (e.g., cycling TBV's during a loss of ICS power i situation) at the simulator is not as great as what would be seen at the plant. Operators should expect a greater effect on
~ the primary side if cycling steam valves at the plant. , Maintaining a narrower secondary pressure band is one way of minimizing this greater response at the plant. ~
- 6. The E0P's direct the operator to isolate the OTSG's (one or both) under certain circumstances and provide a list of valves i
- that need to be closed to accomplish the isolation. The i following valves are not modeled at the simulator, but must be l closed at the plant in order to isolate an OTSG
i
- Main steam to AFH pump ,
} Pain steam to reheaters
- Main steam to pegging steam (' A' OTSG only)
! - OTSG blowdown l I B-9 4 i _ , - _ _ , - _ _ _ . _ - - _ _ _ _ , _ , . _ - . ,_ _ _ _~
- 7. The AFH SFAS valve controls at the simulator are located on the operator's console rather than the SFAS panels as at the plant.
This places the burden of operation for these valves on the console operator who is normally not responsible for operating these valves.
- 8. For HPI tooling, the E0P's direct the operator to initiate SFAS channels IA & IB with their respective push buttons. To get the same equipment response at the simulator, the channel 1/2 and 5/6 initiation push buttons must be used.
- 9. If the operator has to open the 38 and 3C2 breakers to drop the rods at the simulator, the CCH pumps, air compressors and PCH pumps have to be restarted after the breakers have been reclosed.
Also the breakers are located differently on the electric control panel than at the plant.
- 10. Certain steps within the E0Ps direct the operators to re-open the seal return isolation valves following their automatic closure on an SFAS. At the simulator these valves do not auto-close on SFAS.
- 11. The simulator " Loss of ICS Power" transient differs from that at the plant in the following ways:
- a. There are no auto-close switches for the TBV's and ADV's on the operator's console. The " circuitry" for these switches is modeled, however, the simulator instructor must actuate from the simulator control console.
- b. There are no AFH valve controllers on H2PS. The AFH Bailey valves go to their designed pre-set position. However, manual control remains on the operator's console Bailey stations.
l l c. The power history associated with the transient is ! relatively high (may be design decay heat level). Thus, I the transient is essentially an undercooling transient l untti the AFH valves are opened further. 1
- d. There is no "ICS Power / Fan failure alarm" at the simulator,
- e. The reducer control failure for the auxiliary steam system is not modeled at the simulator. At the plant, this will l fall to the 50 percent position. The auxiliary steam MOV's on HISS are also not modeled at the simulator.
FINDING - B.8: Recognition of ICS Power Condition
- 8. The operators who investigated the loss of ICS power did not adequately understand the ICS power system configuration. When i 120 Vac power is still available from the IC bus and the ICS de l power supplies de-energized, the most credible cause for the l
B-10
I loss of ICS de power was the opening of switches S1 and S2. However, the operators did not recognize this fact and, as a O result, did not- shut the switches until 26 minutes into the
-transient. The fact that several operators did not recognize that switches Si and_S2 were OFF suggests that their training on .
this crucial system was not adequate. In addition, although simpilfled drawings of the non-nuclear instrumentation.(NNI) power supplies were posted on the NNI cabinets, comparable drawings for the ICS power supply had not been provided. DISTRICT RESPONSES , Since the event training has been conducted on the design and operation of the ICS/NNI power supplies. In addition, Si and 52 labeling has been engineered and installed at the switch location. A one-line ICS power supply diagram has been posted on the cabinet door to aid the operator in troubleshooting power supply problems. These schematics are similar to the previously installed schematics on the NNI cabinets, and are made of durable engraved plastic laminate. A specific causality procedure for the ICS has been developed and is referenced by the Emergency Operations Procedures for use once the operator has established a stable plant condition, following loss of ICS power. FINDING - 5.9: Damaged Hand Operator on AFW Valve -
- 9. It does not appear that non11 censed operators properly operated the AFW (ICS) flow control valves. An operator applied O excessive force with a valve wrench to close an AFN (ICS) flow control valve. He did so because he had not accurately determined the position of the valve while attempting to shut it completely. As a result of his actions, the valve was damaged, reopened, and the manual (local) capability to operate the valve was lost. These consequences suggest training weaknesses in the acceptable use of valve wrenches, the proper methods for manually operating and overriding ale-eperated valves, and the use of available and back-up indications to determine valve positions. These weaknesses suggest areas where hands-on training rather than walk-through or talk-through training may be necessary.
DISTRICT RESPONSE A specific policy preventing the use of the valve wrenches on gear-drive valve operators hai been estabIlshed. Formal training on this policy and on the proper operation of such valves has been provided to the operators. Tlie training and qualification requirements for operators have been changed to require that each individual operate certain valves, such as the Aux 111ary Feedwater Control valves, so that they are famillar with the feel and characteristics of the valve. Included in the training and operating policies are the requirement to utilize the available indications of valve position and the need to communicate with the control room. O B-11
/
The Operations Training Program has been revised to include: 1) Extensive hands-on OJT in the manual operation of the various types of handwheel-controlled valves expected to be encountered, and 2) detailed classroom training on the normal and manual operation of the TBV's, ADV's, AFH valves and various limitorque valve configurations. FINDING - 8.10: Radiological Controls and Emergency Preparedness
- 10. While the deficiencies in SHUD's radiological control and emergency preparedness programs during the December 26, 1985 incident did not jeopardize the public health and safety due to the relatively minor radiological consequences of this incident, they do indicate weaknesses in SMUD's program and the training of Rancho Seco personnel.
DISTRICT RESPONSES The Olstrict has undertaken a program to significantly enhance the Emergency Plan and the effectiveness of its implementation. Meetings have been held with federal, state, and county representatives to resolve details of procedures and hardware configuration which were identified as impediments to effective implementation. The state and county voice notification system is being upgraded. The December 26 event highlighted communications and procedure adherence issues. Proactive steps have been taken by the Olstrict to get combined training between the Rancho Seco communicators and their counterparts in the counties. Site visits, both ways, have occurred and are now a part of the program. The Emergency Plan itself has been revised to improve and simpilfy its use during emergency conditions. The plant operations staff has received training and simulator practice on effective command and control of complex events. Management policy is clearly stated that the Shift Supervisor has overall plant responsibility and is primarily responsible for the effective implementation of the Emergency Plan. Plant and Emergency Plan duties are clearly stated and preassigned within the operating crew. The Shift Supervisor / Emergency Coordinator's role is to provide the overview and direction to the crew. He is not to perform as a control room panel operator. The Emergency Planning Department conducts frequent site exercises to provide practice training in the use and implementation of the Emergency Plan. In addition, extensive Emergency Plan usage is included in simulator drill training conducted for all licensed operators. Offsite and Onsite Emergency Response Organization (ERO) Annual Training has been completed and records are available in the Nuclear Training Department (NTD) Library. A longer-term project is under way which intends to simplify the Emergency Plan in terms of making it more " user friendly". A major benefit is expected which will be its ability to successfully mitigate complex scenarlos which involve multiple casualties such as, radiation release coincidence with personnel injury and a fire. 0-12
The Emergency Plan is being revised to include the on-site Radiation Protection (RP) technician as part of the Emergency Plan. O Improvements needed in the Radiological Controls, as observed in the December 26. 1985 event, are addressed in Finding 20. FINDING - 8.11: Installation of EFIC
- 11. The NRC staff was led to believe that the Emergency Feedwater Initiation and Control (EFIC) system would be installed in 1984 in response to a number of NRC requirements, including TMI Action Item II.E.1.2. Apparently SMUD decided to install an alternate system in response to II.E.1.2. SMUD's intent to satisfy II.E.1.2 with this alternate design was not made clear to the NRC staff, was not approved by the staff, and may not have complied with the requirements of II.E.1.2. As a result, the EFIC system, some features of which would have reduced the
{ severity of the December 26, 1985 incident, has not yet been j installed at Rancho Seco. DISTRICT RESPONSE The AFW/EFIC (II.E.1.2) scope and schedule changes had been provided to the NRC staff. The NRC issued Safety Evaluation Reports in January and September 1982, assuming EFIC installation. On October 22, 1982, the District indicated that it would install interim safety grade AFW modifications and that EFIC was separate and beyond the AFW upgrade requirements of NUREG-0737. The District indicated at that time that EFIC would be installed by Cycle 7. This schedule was confirmed by the District on December 14, 1982, at which time the Cycle 7 outage was expected to occur in the fall of 1984. The schedule for the interim safety-grade modifications was specified in a confirmatory order dated March 14, 1983. On April 28, 1983, the District submitted a revised AFW system description describing the interim AFW upgrades. NRC confirmed their understanding in a SER on the status of the AFW system dated September 26, 1983. Then in a series of Ilving schedule submittals, the District informed the NRC that the EFIC installation was scheduled in two phases (Cycle 8 and Cycle 9). This approach was understood, and acknowledged, by the NNR staff during a meeting in October 1985, at which time, the District committed to accelerate the EFIC installation schedule to accomplish as much as possible during the Cycle 8 outage, with the balance of the installation to be completed during the Cycle 9 outage. While it has been the District's position in the past that provision ' of safety-grade auto start of AFW via the SFAS was sufficient to meet the intent of NUREG 0737 ltem II.E.1.2, the point is no longer O germane due to the decision to immediately install EFIC and related 5-13
AFH upgrades. The upgraded AFH with EFIC meets fully the requirements of NUREG 0737 item II.E.1.2. This is underscored by the SER received for this concept in April 1983. { } The EFIC System goes substantially beyond NUREG 0737 requirements. The primary purpose is to provide:
- a. Automatic initiation and control of AFH.
- b. AFH indication with a safety-related Class IE instrumentation system that is independent of the ICS and NNI.
The EFIC system also fulfills several secondary purposes which provide:
- a. One train of AFH that is diversely powered and independent of AC Power.
- b. Level control of the OTSG's through automatic and reliable AFH flow control.
- c. Better control of paths through the ADVs of excessive steam flow alleviating the problem for OSTGs to boil dry,
- d. ASTG overftll protection,
- e. Redundant automatic MFH isolation on high OSTG 1evel and low OSTG pressure, q FINDING - 8.12: Reactor Vessel Thermal Shock
- 12. Although the RCS temperature dropped 180'F in 26 minutes, it would have had to rapidly drop another 215'F (i.e., to an RCS temperature of about 170*F) while pressure was maintained at approximately 1,400 psig, in order to seriously threaten reactor vessel integrity.
DISTRICT RESPONSE The District agrees with this finding, based on calculations provided by B&H. In its evaluation for the District dated February 1986 B&W also calculated the effect of the December 26, 1985 evcnt on the reactor vessel as a function of the number of cooldowns consumed in l the transient and the cumulative total of cooldowns versus the number designed into the reactor vessel. The B&W evaluation of the effects of the December 26,1985 event concluded that the event consumed "0.3 cycles" of the 240 designed.
- . To date, all overcooling and abnormal events in combination have l consumed five cycles based upon NSS fatigue analysis. Normal cycles
! have totaled less than 35, leaving 200 heat-up/cooldown cycles I available. This number provides sufficient margin for achieving the balance of the plant design life. B-14
The Electric Power Research Institute (EPRI) applied a new nonmandatory ASME Code Section XI, Appendix XX, to the December 26, O 1985 event and its effects upon the Rancho Seco Reactor Vessel. The ASME evaluation procedure allows demonstration of adequate structural integrity of the reactor vessel-beltline without doing further integrity analyses as long as the reactor coolant pressure hasnotexceededdesignpressure(2,500 RTNOTS has psi)andTclationsshowthat not been less t5an 55 F during the transient. Calcu the Rancho Seco December 26, 1985 transient met these criteria. The reactor vessel calculations demonstrate that the Rancho Seco reactor vessel beltline region has adequate structural integrity for return to service without further evaluation. FINDING - 8.13: Reactor Vessel Integrity
- 13. The December 26, 1985 overcooling incident does not appear to have seriously threatened the integrity of the Rancho Seco reactor vessel. However, the plant has had a number of overcooling incidents in its 12-year operating history. Each time this occurs, the potential exists for additional operator errors and equipment failures that might have exacerbated (sic) the event and seriously threatened reactor integrity. Thus, the significance of this incident Iles in the fact that under alternate scenarlos, more serious consequences could occur.
DISTRICT RESPONSE The issue of reactor vessel integrity was discussed in the response to Finding 12. The District agrees with the IIT that the December 26, 1985 event does not appear to have threatened reactor vessel integrity and that the District's programs should focus on eliminating the precursor events which provide the situations which can lead to serious events. The Olstrict's Plant Performance and Management Improvement Program specifically addresses investigations which are of a retrospective nature and focus on preventing trips and thereby avoid challenging operators and/or safety systems. In this way we prevent a transient that could cause the post-trip response to leave acceptable pressure and tesperature limits and begin to develop the characteristics of a serious or challenging event. FIN 0!NG - 8.14: Use of ICS in FSAR Deslan Basis Events 14, It is not clear that the overcooling transient was within the Final Safety Analysis Report (FSAR) analysis of the Rancho Seco plant. Although PTS has been addressed generically, the FSAR accident analysts for Rancho Seco does not address this issue. The most comparable analysts in the FSAR is for the cooldown due to a main steam line break. However, this analysis included only 100 seconds of the transient. In addition, the Rancho Seco 8-15
FSAR analysis of main steam line breaks appears to be flawed and nonconservative in that it assumes that the nonsafety-related ICS operates successfully to mitigate the consequences of the accident. DISTRICT RESPONSE The Rancho Seco original FSAR description of the main steam line break (HSLB) consisted of two analyses: MSLB with ICS actions, and MSLB without ICS or operator actions The HSLB analysis with ICS actions is conservative with respect to maximizing off-site doses. The analysis assumes I percent failed fuel with the technical specification steam generator tube leakage. The ICS actions are assumed to occur to maximize the cooldown time to decay heat removal system operation, thus maximizing the releases via the intact steam generator. The HSLB analysis without ICS or operator action is conservative with respect to maximizing the potential for a return to criticality and potential adverse effects on the fuel. During the licensing phase of Rancho Seco, the issue of HSLB inside the reactor building was raised. This concern was addressed by installation of automatic feedwater isolation, performed by the Main Steam Failure Logic (HSFL). The HSFL is independent of the ICS, consists of redundant actuation channels and is battery backed; however, it is not safety grade. Reactor Building Containment integrity is not required for the HSLB, as the worst case for dose considerations is the HSLB outside the reactor building. The above design basis was clear in the original FSAR. The analysis for HSLB without ICS, or operator action, was contained in a response to an NRC question. During the compilation of the Updated Safety Analysis Report, this analysis was poorly worded when incorporated into the text. The separation of the "with and without ICS" analyses is not clear, and can lead to incorrect conclusions. The Olstrict is currently revising the description of the HSLB analyses in the USAR for clarity. This clarification was included in the USAR update submitted in July 1986. The cooldown rate of the HSLB analysis, without ICS or operator action, bounded that of the December 26, 1985 event. The cooldown l rate of the analysis was such that high pressure injection (HPI) was I initiated in 23 seconds, followed by core flood tank (CFT) injection 47 seconds after event initiation. During the December 26, 1985 event, HPI/SFAS initiation occurred after three minutes and the pressure never reduced to that needed for CFT injection. A review has been made of the other design basis accident analyses in Chapter 14 of the USAR and it has been determined that ICS, or other 8-16
nonsafety-grade equipment action is not assumed in the mitigation of those accidents, with the exception of fuel handling accidents. For the fuel handling accident, the releases are assumed to be filtered (V, ,) through the auxillary building filters, which are nonsafety grade. Credit for these filters is appropriate as they are subject to technical specifications and the system must be operating during fuel handling operations. This design basis was clearly described and reviewed in the final Safety Analysis Report. The subject of PTS is addressed in Chapter 4 of the Rancho Seco ' USAR. The Babcox and Wilcox report BAW-1791, "B&W Owners Group Probabilistic Evaluation of Pressurized Thermal Shock - Phase 1 Report," June 1983, is referenced and described. Numerous PTS events are evaluated in BAH 1791 including events similar to the December 26, 1985 event. BAW 1791 is discussed furtaer under Finding 26. With the PTS rulemaking in December 1985, the NRC estM:11shed screening criteria for PTS. The District's response t) the screening criteria has been submitted, and will be incorporated into the USAR in the 1986 update scheduled for July 1987 submittal. FINDING - 8.15: Precursors to 12-26 Event
- 15. There were a number of precursors to the December 26, 1985 incident at Rancho Seco. These precursors indicate that improvements in the reliability of the ICS and procedures to efficiently mitigato a loss of ICS power have not been developed or implemented at Rancho Seco despite numerous efforts on the part of the NRC staff to improve the reliability of the ICS and to ensure that the necessary procedures to efficiently mitigate such an event would be available to the operators. While the staff had raised these issues on a number of occasions over the past six to eight years SMUD personnel had not implemented the actions, and the NRC staff had not taken effective action to ensure that the improvements in reliability and the procedures were developed and implemented at Rancho Seco. The specific findings associated with these precursors include: (specific responses follow.)
O! STRICT RESPONSE The District is committed to a permanent Precursor Review program. The Precursor Review, currently under way as part of the Plant Performance Improvement Program, is resulting in the identification of recommendations not only to address specific precursors, but also to determine whether the District's previous analyses of precursors was too narrow in scope and, therefore, worthy of additional action. O B-17
FINDING B.15.a: Power Supplies to ICS/NNI
- a. Although the ICS power supply is similar to the NNI power supply, particularly with respect to the role of the power supply monitor, SMUD's principal emphasis following the lightbulb incident in March 1978 was on the NNI rather than on the ICS. This emphasis seems to have biased SMUD's subsequent reviews of issues associated with the NNI and ICS.
DISTRICT RESPONSE The reactor core remained protected throughout this event. Following - this event, SMUD was concerned that this previously unrehearsed situation had caused considerable uncertainty with respect to the validity of the instrumentation in the Control Room. As a result, the Olstrict conducted an extensive review and provided a wide range of modifications and procedural enhancements, mostly related to NNI. The upgrades to the ICS power supply and power supply monitor are discussed under Finding 1. FINDING B.15.b: January 5. 1979. loss of ICS Power
- b. The loss of ICS power transient at Rancho Seco on January 5, 1979, was similar to the December 26, 1985 incident. However, it was not as severe as the "lightbulb incident" and did not receive the same level of attention. As a result, changes in the design of the ICS were not made and procedures for loss of ICS were not developed.
DISTRICT RESPONSE The course and consequences of the January 5, 1979 loss of ICS power event were very similar to the December 26, 1985 event. In both cases, the reactor core remained protected throughout the event. SFAS actuated on low RCS pressure to maintain RCS inventory core subcooling. The upgrades to the ICS power supply and power supply monitor are discussed under Finding 1. FINDING B.15.c: ICS Reliability Study. BAW-1564
- c. In March 1979, B&W issued a report (BAW-1564) in which they analyzed the reliability of the ICS. Although the B&W analysis noted a number of changes that appeared to be warranted in the ICS, SMUD concluded that no changes were necessary. A subsequent analysis of the ICS by the Oak Ridge National Laboratory criticized the B&W analysis and noted that it was of limited scope and did not appear to meet the requirements of the original order. The NRC staff concluded that no immediate changes were required at Rancho Seco as a result of the B&W analysis. The long-term issues associated with the B&W report B-18
I l were to be considered in Unresolved Safety Issue (USI) A-47,
" Safety Implications of Control Systems." !
i v DISTRICT RESPONSE ! }! The following clearly establishes the status of 8AW-1564. The six '
; recommendations from the report are being input to the evaluation i process described in Section 4A. The present status of each is j discussed at the end of the following except from the ASLB decision. l j The ICS reliability study, BAW-1564, was litigated as to its i i completeness and adequacy before the Atomic Safety and Licensing ,
i Roard which issued a decision (L8P-81-12) dated May 15, 1981. The ! l section of that dectslon pertaining to the ICS 1s provided below. ! II. FINDINGS OF FACT i , i l A. Integrated Control System : 2
- 18. Board Question H-C 16: ,
! Is the failure mode and effects analysis for the l Rancho Seco integrated control system complete and l l adequate? l l One of the long-term actions directed by the
! Commission in its order of May 7, 1979, was that i
"[tlhe licensee will submit a failure mode and effects !
analysis of the Integrated Control System to the NRC l Staff as soon as practicable." 44 Red. Reg. at 27779 l (1979). Such an analysis was performed by 88,W for , Licensee as part of 8&W's study of the reliability of f , the integrated control system ("ICS"). The results of B&W's reliability study are contained in B&W Report i BAW 1564, " Integrated Control System Reitability l Analysis." CEC Ex. 3.
- 19. In order to assess the completeness and adequacy of
- B&W's analysis, it is important, first, to understand l the Rancho Seco ICS and the staff's concerns regarding -
i it. The ICS 1s an automat 1e control system whose j j basic function is to continuously match the unit's J power generation to its load demand. The ICS does I this by coordinating the rate of steam generation and ! ] the steam flow to the turbine. NRC Staff Testimony of l Dale F. Thatcher Relative to the Integrated Control l System (Board Question 16), following Tr.1163 j (" Thatcher ICS Testimony"), at 2. f 20. During normal operations, the ICS provides proper ! l coordination of the reactor, steam generator, ! 7 feedwater control, and turbine. Proper coordination ! consists of producing the best load response to unit B-19 1
load demand within the limitations and capabilities of the plant equipment. Id. at 3.
- 21. The ICS includes four subsystems: unit load demand control, integrated master control, steam generator control, and reactor control. Id. at 2. Each of these subsystems (except for the unit load demand control) regulates and interacts with a number of other plant control systems, such as the control rod drive system and the feedwater pump and valve controls. Id. at 3. The ICS can maintain a constant average reactor coolant temperature at power levels between 15 percent and 100 percent of load and can maintain constant steam pressure at all loads. Id. at
- 3. During load changes and system upsets, the ICS applies signals to control major parameters (feedwater flow, steam pressure, reactor power, and reactor coolant temperature) in such a manner as to achieve optimum overall plant response without challenging the safety systems. Testimony of B. A. Karrasch and R. C.
Jones, fol. Tr. 535 ("Karrasch-Jones") at 7-9. It has been demonstrated that the ICS can reduce power from 100 percent to 15 percent and maintain that level should the turbine trip without calling upon the reactor's protective systems (Karrasch-Jones testimony at 10), although presently an anticipatory reactor trip on turbine trip has been added so that the ICS can no longer perform this function. Id. The ICS was thus designed to keep the reactor on line during off-normal conditions and enhance plant availability. Id. at 7; Tr. 1076. If, because of protective system actions, the reactor does shut down, the ICS will control steam pressure and maintain a preset steam generator level by controlling steam and feedwater, so long as either main or auxillary feedwater is available. Tr. 1105, 1118, 1119.
- 22. CEC has emphasized, both in its cross-examination and in its Proposed Findings, the notion that it is the sensitivity of the B&W steam supply system to suondary side conditions which makes the ICS necessary and which, therefore, makes reliability of the ICS a very important matter. CEC Proposed Findings at 31-32; Tr 1103-1105. Both Staff and Licensee emphasize the similarity of the ICS to the systems used at other power plants, including fossil-fueled plants. Staff's Proposed Findings at 11; Licensee's Proposed Findings at 24; Karrasch-Jones Testimony at 7. It appears that, in the days shortly after the THI-2 accident, the Staff was concerned that the ICS could cause or contribute to an incident.
Thatcher ICS Testimony at 5; CEC Ex. 26 at 1-5, 2-9. In particular, the Staff then believed that an ICS 8-20
malfunction could prevent auxillary feedwater (AFW) from being supplied during a loss-of-main-feedwater ! [mi transient or could cause such a transient. Id.; Tr. V 1270-72. l 23. The first concern was addressed on a short-term basis i in the Commission Order of May 7, 1979, by requiring Licensee to "[d] develop and implement operating ! procedures for initiating and controlling auxillary feedwater independent of Integrated Control System control." 44 Fed. Reg. at 27779 (1979). The adequacy of Licensee's compilance with this aspect of the May 7, 1979, order was established by the Staff by i visiting the site and conducting examinations of the l operators to verify the adequacy of their training. This evaluation included a walk-through of some of the procedural aspects of manually controlling AFW Independent of the ICS and a review of plant diagrams to verify that the valves that would be utilized for AFW flow control were indeed independent of the ICS. Thatcher ICS Testimony at 4, 5; Tr. 1386, 3730, 3731; Staff Evaluation at 13,
- 24. A permanent solution to the first concern has been provided by Licensee's safety-grade AFW control system independent of the ICS. This modification will completely remove the operation of the AFW system from i A the ICS. Thatcher ICS Testimony at 51 Tr. 1273.
Ih
- 25. It was the second concern relating to the ICS that led l the Staff to ask that a failure mode and effects
- analysis ("FMEA") of the ICS be performed. Since the i Staff was interested in the potential role of the ICS l as the instigator of a transient, it sought to have an l analysis made of the reliability of the ICS and the l effects of failures of that system on the plant's l operation. Tr. 648, 937-39; Tr. 1270-73. A FMEA is a i systematic procedure for identifying the modes of failure of a system and for identifying their consequences. It seeks to determine if any single failure in a system can prevent the system's function. It is considered to be the first general step of a reliability analysis. Thatcher ICS Testimony at 6. Accordingly, an ICS FMEA was one of the long-term actions directed by the Commission in its Order of May 7, 1979. As a long-term action it was not a condition of restart.
- 26. B&W performed the FMEA as part of its reliability analysis of the ICS. It determined the expected effects upon the B&W steam system from single failures of ICS inputs, outputs, and internal modules. The Rancho Seco plant was chosen specifically as a O
B-21
representative design for all the B&W units for the analysis. The analysis was complemented with an evaluation of field data from all B&W operating plants and a computer simulation to confirm the effects of various ICS failures on r.ssociated equipment. Karrasch-Jones testimony at 11; Staff Ex. 5 at 3. The analysis was made a part of our record as CEC Exhibit 3, " Integrated Control System Reliability Analysis," BAW--1564, August 1979, as was a review by Oak Ridge National Laboratory of the analysis (Board Exhibit 1). Also a part of the record is Staff Exhibit 5, the Staff review of both reports.
- 27. Fundamentally, B&W's analysis of the reliability of the ICS thus consisted of three parts: the FMEA, a computer simulation used to study the effects of failures in more detall (both of these specific to Rancho Seco), and a review of operating experience from all B&W operating plants. Board Ex. I at 5.
- 28. The overall conclusion of the FMEA was that the reactor core remains protected throughout any of the ICS failures studied. For those postulated ICS failures which could cause reactor trip, the safety systems would operate Independently of the ICS malfunction and they were assumed to operate properly. The overall conclusion from the operating experience evaluation was that ICS hardware performance has not led to a significant number of reactor trips. It was, in fact, concluded that the ICS has prevented more reactor trips than it has caused and, accordingly, its not effect has been a reduction in the number of challenges to the Reactor Protection System. It was further concluded that the FMEA shows that no ICS failure can prevent proper safety system functioning and that the operattrg expertence demonstrates that the ICS is a reliable system with regard to preventing plant upsets.
Karrasch-Jones Testimony at 11-12.
- 29. The ORNL Review concluded that although the ICS and related control systems contain areas which can be potentially improved, the ICS itself has proven to have a low failure rate and it does not appear to precipitate a significant number of plant upsets.
Specifically, the examination of the failure statistics revealed that only a small number of ICS malfunctions resulted in a reactor trip (approximately 6 of 162). In its review, the ORNL concluded that the ICS is a "significant asset to plant safety and availability." Board Ex. I at 11. O B-22
- 30. While agreeing with 84W's findings and conclusions and with the recommendations made by B&W for further O leprovements in areas relating to the ICS. the ORNL Review pointed out a number of perceived deficiencies in B&W's approach to the FMEA portion of the reliability analysis. Tr. 1706-07, 1774. Board Ex. I !
passim. The main criticism leveled at the FMEA by l ORNL was that the scope of the FMEA was too limited, ; leading to results having only limited value. Board Ex. I at 4. The scope limitations identified by ORNL < were: (1) not considering the interactions between j plant safety and nonsafety systems such as ICS; (2) , not including analysts of failures of plant systems l external to the ICS: (3) not considering multiple ; system failures; and (4) uttitration of functional versus component diagrams as the building blocks in l the analysis. Board Ex. I at 3, 4, and 6 through 8.
- 31. It was, indeed, critical language from Board Exhibit 1 l that formed the basis for this Board's inclusion of ;
BQNC 16 in this hearing. In particular, such l statements as:
...the B&W analysis is more notable for what it does not include than for what it does include."
and "
...Because of this limited scope, the results f are of Ilmited value."
(Board Ex. I at 3 and 4) l would surely give one pause if taken out of context. t We note, however, the following points about each of , the four numbered limitations of scope set forth above: Point (1): Interactions between safety and nonsafety systems such as ICS were not ! considered. That is true, but such analysis was not specifically required by the NRC's M4y 7 ' 1979, order. A study of such actions is under way for all plants as a part of the Staff's
" Integrated Reliability Evaluation Program" (IREP) which has as one of its objectives to identify the risk significance of systems Interactions originating in the ICS of S&W plants. Thatcher !CS Testimony at 8.
Point (2): Fallures in systems external to the ICS were not included. This is beyond the scope of the May 7 order. Actually, the B&W analysts did include some such failures in that it included failures in the inputs to and outputs from the ICS. Tr. 641-83, 1083-86. O B-23
Point (3): Multiple failures were not considered. They were not, nor is it usual to include multiple failures in a FMEA. Tr.1083; Thatcher ICS Testimony at 6-7. Such an analysis is usually used to determine whether a single failure can prevent operation of a safety system. Id. The ICS has not been required to meet the single failure criterton and was not previously analyzed; such analysis can, however, be used to identify failure modes which lead to undesirable consequences. Id. at 7. As we noted above, no such consequences were found. Point (4): Functicral block diagrams were used rather than component diagrams to analyze the ICS. By this we mean that only the general functions of the ICS were used and failures of each functional block were considered, rather than identifying each specific piece of equipment and considering its failure. Board Ex. I at 6,
- 10. It is possible that presently undisclosed interactions between functions might be revealed by examining specific component failures. Board (
Ex. I at 6. It is also possible that (if the failure rates of specific components were known) one might estimate the probabilities of various modes of failure by that method. Tr. 1086. However, by taking the approach which they took, the B&W analysts clearly met the requirements put upon them. Further, it is not clear to the Board that a component-based analysis and estimated failure rates would give a clearer picture of reliability than the " actual history" approach which B&W supplied in addition to the FHEA. He think, in fact, that the reverse is true.
- 32. He note that the first conclusion of the B&W analysis was that:
- 1. The (Non-Nuclear Instrumentation) power sources (external to ICS cabinets) have been vulnerable to single failures and human errors that have led to reactor trips and plant overcooling. (CEC Ex.
3 at 2-2) and we note further that it was failure of the Non-Nuclear Instrumentation power supplies that initiated the incident. 3717-18.
- 33. In other areas identifled by the study, t.lcensee is considering changes to increase the reliability of the reactor coolant flow input signal to the ICS (Tr.
3703-04), and has developed procedures to improve the
" tuning" of the ICS to the balance of the plant, B-24
- -_-_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ A
having trained operators further in ICS control. Tr. l 3704-05.
- 34. Thus, the ICS itself is even better now than it was when the B&W analysis was performed. As to what that analyses showed, even Board Ex. 1, which was, as we noted, in some respects critical, says:
The manufacturer contends, and we agree, that (1) the system prevents or mitigates more upsets than it causes and (2) the system is generally superior to manual or fragmented control schemes. Board Ex. 1 at 15.
- 35. In sum we find that the FMEA was undertaken in response to certain Staff concerns, that the results of the analysis should allay those concerns, and that the FMEA was adequate and complete for its purpose. I We note that it raised other issues whose resolution l would be expected to yleid an even more reliable and I safer plant (para. 33 supra), and that those issues are being acted upon. Although the need to perform a broader study of the B&W control system and its role in the initiation and the mitigation of transients has been identified and it will be carried out in the IREP, we see no reason to believe that the Rancho Seco plant would present a hazard to public health and safety during the ongoing investigations and upgrading.
The District and the B&W Owners Group have embarked on a program to perform a comprehensive ICS evaluation. A portion of that evaluation will be a new ICS FMEA, which will supercede the existing effort reported in BAN 1564. B&W report BAH 1564 identified six recommendations for further investigation. The status of the District's actions with respect to those recommendations is provided below: RECOMMENDATION 1 8&W recommends that the NNI/ICS power supply be reviewed for possible changes to enhance reliability and safety. DISTRICT RESPONSE The NNI/ICS power supplies have been the subject of several reviews and subsequent modifications to enhance reliability. As described earlier under Finding 1, the NNI/ICS now have their stand-by source of 120 ac power coming from dedicated diesel-backed inverters. In addition, new S1/S2 switches have been installed and surveillance / preventative maintenance done on the power supply /distributton circuits. The root cause of the December 26, , 1985 event was a bad " crimp" on a factory-installed ICS power O B-25
distribution wire. Over 40,000 terminations have been subsequently inspected / tested / upgraded in cabinets supplied by that manufacturer. While a generic problem was not found, the effort should significantly improve confidence in the hardware reliability. RECOMMENDATION 2 B&W recommends that the input signals from the NI/RPS system to the ICS - specifically the RC flow signal - be reviewed for possible changes to enhance reliability and safety. DISTRICT RESPONSE
.J This recommendation is being implemented prior to restart. In addition, the District is participating in the extensive ongoing BH0G program to review the other ICS inputs for ways to improve their reliability. This effort is a long-range program and will not be complete by restart.
RECOMMENDATION 3 B&W recommends that the ICS/ BOP system tuning, particularly feedwater condensate systems and the ICS controls be reviewed for possible changes to enhance reliability and safety. DISTRICT RESPONSE The District had reviewed the ICS/B0P system tuning and had performed ICS tuning prior to the December 26, 1985 event. The ICS was performing in a well tuned fashion at the time of the event. ICS tuning will also be performed during Restart testing. ICS tuning is a part of the Olstrict's ICS maintenance program. RECOMMENDATION 4 , l B&W recommends that main feedwater pump turbine drive minimum speed control be reviewed for possible changes to prevent loss of main feedwater or indication of main feedwater to enhance reliability and safety. l l DISTRICT RESPONSE This recommendation is being addressed in the System Status Report for the Main Feedwater System. With the installation of EFIC, it is I no longer necessary to use MFP discharge pressure to initiate AFH l pumps, hence, the minimum speed can be set based upon operational requirements. RECOMMENDATION 5 ! B&W recommends that a means to prevent or mitigate the j consequences of a stuck-open main feedwater start-up valve be reviewed to enhance reliability and safety. B-26 l
DISTRICT RESPONSE O Both the Main and Start-up Feedwater Control Valves are being provided with Class 1. Bottled Instrument Air to ensure motive power for closing. Furthermore, EFIC included the installation and control of a new motor-operated valve downstream of the Main and Start-up Feedwater valves which will give Class 1 1 solation capability for feedwater. This satisfies this recommendation. RECOMMENDATION 6 B&W recommends that a means to prevent or mitigate the consequences of a stuck-open turbine bypass valve be reviewed to enhance reliability and safety. DISTRICT RESPONSE Although Class 1 Bottled Instrument Air is being provided to the TBVs to ensure motive power, the possibility of being mechanically stuck open remains. Therefore, a motor-operated isolation valve is being installed upstream of the TBVs to ensure a timely remote isolation capability. FINDING B.15.d: District Response to IE Bulletin 79-27
- d. As a result of th' loss e of power to NNI and ICS at Oconee in November 1979, NRC issued Bulletin 79-27 describing a number of actions to be carried out by licensees. Although the Bulletin C raised significant concerns about the consequences of a loss of power to ins.trumentation and control systems, SMUD concluded that no additional design modifications were necessary and that event-oriented procedures to deal with such events were not necessary. It would appear that Bulletin 79-27 was initially intended to solicit detailed information from licensees that could form the basis for an in-depth review of the issues associated with control systems comparable to the review of safety-related systems conducted as part of an operating license review. Based on the initial scope of the review, the conclusion was reached that SMUO's response did not contain sufficient information and did not adequately address the concerns in the Bulletin. After the progressive narrowing of the scope of the review, it was decided that the SMUD response was adequate, despite what appear to be a number of weaknesses in the SMUD response. Thus, the conclusion was finally reached that SMUD had provided reasonable assurance that they had addressed the concerns in Bulletin 79-27, and that the long-term implications of Bulletin 79-27 would be addressed as part of USI A-47.
DISTRICT RESPONSE The District is revisiting Bulletin 79-27 as a part of the precursor review program to ensure that the concerns identified have been O adequately addressed in a broad scope fashion. B-27
The Precursor Review Group reviewed the District's original response to IE Bulletin 79-27 and determined that the response was ao longer appropriate and that further technical review was required. The Deterministic Failure Consequence Analysis Group reviewed the bulletin for identification of instrumentation and failure consequences. The District is performing a comprehensive review of compilance to IEB 79-27 to determine necessary procedural and hardware enhancement. The District will be in compliance with IEB 79-27 at restart. I FINDING B.15.e: District Response to February 1980 Loss of NNI at Crystal River
- e. Following the February 1980 loss of NNI power at Crystal River, the NRC identified an issue about the failure mode of atmospheric dump valves (ADV) on loss of ICS power. SMUD's response to this issue did not include the other valves at Rancho Seco that repositioned on loss of ICS power (i.e., they confined it to the narrow issue associated with the ADV's). In addition, SMUD deferred this narrow issue to installation of the EFIC system, which to date has not been installed at Rancho Seco. The NRC found this response to be acceptable.
DISTRICT RESPONSE Following the December 26, 1985 overcooling event, the District ' performed modifications to the controls of the atmospheric dump valves, turbine bypass valves (TBV), and the auxiliary feedwater (AFH) flow control valves. These modifications are discussed under Finding 2. FINDING B.15.f: District Response to NUREG-0667 ,
- f. Because of concerns about the transient response of B&W-designed reactors and the role of ICS as an initiator of such transients, NRC conducted an extensive study and made 22 recommendations in NUREG-0667. However, it does not appear that these recommendations were sent to SMUD for action or that the recommendations that are relevant to the December 26, 1985 incident were implemented at Rancho Seco.
DISTRICT RESPONSE The District has implemented many of the NUREG 0667 recommendations. The balance of the recommendations are currently being evaluated by the District and/or the B&W Owners Group as elements of the District's Systems Review and Test Program. O B-28 ! I a
FINDING B.15.q: Significance of Partial Loss of NNI at Rancho Seco March 1934
- g. The March 19, 1984, partial loss of.NNI power at Rancho Seco again demonstrated that the failure of nonsafety-related .
equipment at B&W-designed plants has the potential to cause plant transients and to challenge the operator's capability to mitigate the transient without overcooling and undercooling the primary system. Despite the fact that this event occurred nearly two years ago, the December 26, 1985 incident demonstrates that neither SMUD nor the NRC staff has implemented effective actions to resolve this situation. In questions asked by the staff and responses provided by the B&W Owner's Group following the March 1984 loss of NNI power at Rancho Seco, the Team again sees strong evidence of a narrow focus on the incidents initiated by inappropriate control system actions in response to false inputs from the NNI. The questions in general do not refer directly to the ICS. As a. result, the full significance of the loss of power to the ICS was not addressed. DISTRICT RESPONSE J The District has implemented modifications and is in the process of implementing additional modifications to ensure that the plant will go to a known state with the ability to control decay heat removal independent of the ICS, following a loss of ICS or NNI power. For perspective, the March 1984 event was a partial loss of NNI caused by the effects of the Exciter Hydrogen fire on the class II power O . supply. When that power was lost, the auto-transfer to the vital power supply was interrupted due to a low set point on the over-voltage protection circuit. This occurred approximately one hour following the Reactor shutdown. It resulted in an ADV opening which was immediately recognized and closed by the operator in the ! Control Room using manual ICS control. Approximately two hours later, while troubleshooting the NNI power supplies, there was a short interruption of NNI-X power which did not adversely effect the l plant. FINDING B.15.h: Applicability of " Reference Plant" Studies l f to Rancho Seco
- h. While the scope of the analysis performed under USI A-47 is broad, it appears that to date the actual study includes only those events with the potential to produce consequences outside the design basis of the reference plant. Such events are rare so the study does not appear to address substantive issues of the frequent challenges to protection systems and frequent abnormal operating occurrences, such as those identified in BAW-1564, Bulletin 79-27, and NUREG-0667. In addition, the analysis does not consider the events that are significant at other than the reference plant. Differences in plant design that could cause an event to be significant at another plant are i not adequately considered.
O Therefore, it appears that the B-29
analysis performed to date under USI A-47 does not address the long-term issues raised in bulletin 79-27, BAH-1564, or NUREG-0667 that are relevant to the Dece.:ber 26, 1985 incident. Thus, results of the resolution of USI A-47 are of quite limited applicability to B&W-designed plants beyond the reference plant that was studied. The results are not directly applicable to most other B&W-designed plants such as Rancho Seco because of the differences in the design of the ICS. DISTRICT RESPONSE The resolution of USI A-47 is a NRC staff action. However, the District has implemented a program to identify events that challenge the safety systems or the operators and take appropriate corrective actions. In addition, the B&W Owners Group has embarked on the SPIP Program, which is aimed at reducing the frequency and severity of transients at B&W plants. A portion of this program includes an extensive evaluation of the ICS as a contributor to plant transients. This evaluation will take into consideration the differences in ICS designs between the various B&W plants. FINDING - B.16: Timely Identification of Loss of ICS Power Condition
- 16. It appears that the transient initiator (i.e., the loss of ICS de power) was not fully recognized by control room operators until two minutes after the power was lost. Although the "ICS and Fan Power Failure" alarm alerts operators about ICS power failures, it appears that its importance was somewhat obscured because it also acts as a trouble alarm for fan failure or for loss of one of the redundant ICS dc power supplies, neither of which requires immediate operator actions or initiates a transient.
DISTRICT RESPONSE The operators had determined the loss of ICS power prior to the ', reactor trip, i.e., within the first 15 seconds as a result of the annunciator alarm and the loss of ICS Controller Hand / Auto Indicator lights. Coupled with the immediate "undercooling" effects (due to the Main Feedwater pumps being run back) there was no ambiguity which suggested a " Fan Power Failure." In response to the potential for ambiguous annunciator alarms, the ICS/NNI alarm / annunciators have been completely revised. The key feature of these changes was to , incorporate human engineering concepts into the annunciators. This resulted in two annunciators, one for " trouble", the other for
" failure". This grouping was developed from a review of the required operator response to the various conditions leading to annunciation.
As previously addressed, extensive training and procedural guidance has been developed to enhance operation's capability to specifically identify and diagnose ICS failures. O I B-30
Furthermore, as a long-term effort, a complete reassessment of the Annunciator Systems is to be accomplished to ensure that annunciators g) are provided for all important parameters and that they are (d unambiguous to the operator. This is an element of the CRDR Human Factors committed upgrades which preceded the occurrence of the December 26, 1985 event. FINDING - B.17: Usefulness of Annunciator Procedures Manual
- 17. The Annunciator Procedures Manual was not used by the operators l following the "ICS or Fan Power Failure" alarm. Even if the l Annunciator Procedures Manual had been used, it contained very limited guidance concerning the implications of this alarm and would have been of no value to the operators in recognizing or restoring the loss of ICS dc power.
DISTRICT RESPONSE The Annunciator procedure, along with the other operating procedures (EOPs, cps, SOP's, SP's), have been the subject of a comprehensive Operational Assessment following the event. As a consequence, a number of revisions have been made to incorporate the lessons learned in this event. In addition, a long-term annunciator procedure upgrade program is being implemented which will provide annunciator procedures consistent with the Human Factors /CROR Annunciator upgrades which are also being developed. p The lack of detail in the ICS and NNI Annunciator Procedures has been resolved by referencing the appropriate Casualty or Emergency Procedure and providing more detail to help the operator assess the significance of the alarm. The annunciator procedures are intended to direct the operator to investigate and correct very specific events, not provide direction for overall plant control, which is accomplished by the E0P's complemented by the casualty procedures. Again, the annunciator procedures are to be reviewed and upgraded as part of a long term effort on procedure improvement. FINDING - B.18: Performance of ICS Following Restoration of Power
- 18. The ICS performance upon restoration of power is still not fully understood, especially because performance may depend on the duration of the power interruption. However, when ICS de power is restored, reactor operators regain remote control of plant equipment from the control room. (It is the Team's understanding that the B&W Owners' Group is planning to conduct an investigative program that will include this matter.)
i DISTRICT RESPONSE 1 The B&W Owners Group is evaluating various aspects of the ICS including performance upon restoration of power. This evaluation will consider the results of testing performed at Davis-Besse on the p i ICS performance upon power restoration. B-31
The District will modify the ICS/NNI memory modules to establish a predictable state upon restoration of DC power. The basic approach for ICS or NNI power restoration (following plant stabilization) is to establish plant conditions (through a pre-existing procedure for power restoration) such that there will be little or no effect on the plant regardless of the signal generated by the ICS or NNI during repowering. For example, TBVs would be under control of their independent control circuits, or manually isolated, prior to power restoration. Only when the operator is satisfied that ICS demands are as desired will control be returned to the ICS. Use of the controls has been incorporated into the " Loss of ICS" procedure and training has been performed. Proper functioning of these new controls, and validation of the new procedure for repowering, will be demonstrated during the restart test program. FINDING - B.19: Control Room Indicators Which Fall to Mid-Scale
- 19. Most of the indicators in the control room (both meters and recorders) are part of the NNI system; hence, they are generally independent of the ICS. However, there are exceptions that had not been recognized prior to the December 26, 1985 incident.
For example, the main feedwater (MFH) flow recorders are affected by the ICS. During the December 26, 1985 incident, the recorder failed to a value near mid-scale when MFH flow was actually zero. DISTRICT RESPONSE A detailed review of the ICS drawings was confirmed by testing to observe the effects of Loss of ICS de power. The result was that the following indicators are affected: Main feedwater Flow Recorders, A Loop and B Loop Main Generator Electrical Frequency Error Electrical Frequency Error is a parameter used within the ICS and is for information only to the operator. The Main Feedwater Flow Recorders are affected by the ICS as it is necessary to combine the Start-up and Main flow signals to develop the total feedwater delivered to each steam generator. The mid-scale value which results, following loss of ICS power, has been determined to have no adverse effect upon operations should this event reoccur. The reason is that a loss of ICS power will trip the main feedwater pumps. The indication of Main Feedwater Flow remains high (mid-scale position). The Emergency Operating Procedures (E0P's) direct that the associated pumps then be verified tripped. As an interim measure, the District has labeled all ICS (and NNI) indication. Emergency Operating Procedures and training instruct the operators to use alternate indication in the event of loss of ICS (or NNI) power. In addition, the District will address the issue of mid-scale failure in conjunction with the BH0G I&C Committee. B-32
FINDING - B.20: Adherence to Radiation Protection Requirements m
) 20. Because of a perceived sense of urgency, two nonlicensed
_/ operators made an emergency entry into the make-up pump room without respiratory protection or adequate protective clothing, . neither of which was readily ava',lable. As a result, their clothing was contaminated and they were exposed to airborne radioactivity. DISTRICT RESPONSE Since the event, the appropriate procedures have been extensively revised and training has been completed in response to the lessons learned. Furthermore, management's policies regarding radiation protection and procedure adherence have been clearly stated to ensure that all personnel understand their responsibilities. Additional protective equipment, including respirators, have been staged at locations more convenient to personnel requiring their use to minimize delays during emergency conditions. Significantly, another health physics technician has been added to each shift to provide operations with dedicated Health Physics support. This has been beneficial by improving communications and mutual support between the Operations and Health Physics functions. Deficiencies noted in the content of several Emergency Plan Implementing Procedures (EPIPs) related to radioactivity release alarm setpoints, assessment of offsite dose, and documentation requirements have been corrected. FINDING - B.21: Programmatic Efforts to Disseminate Lessons-Learned and Plant' Changes
- 21. The operators did not remember a recent modification had been made to permit the TBV's and ADV's to be closed from the remote shutdown panel (outside the control room) Independent of the availability of ICS power. This change was made to accommodate a fire in the control room. Although this modification had been incorporated in the control' room fire procedures, SMUD did not review other procedures to determine the applicability of this modification.
DISTRICT RESPONSE The training provided at the time the shutdown panel controls were i installed did address the use of those controls for events other than fire in the control room. That the operators did not remember these controls until after the event was terminated suggests that additional efforts are needed to prevent reoccurrence. Remedial training has been provided on this specific modification. Further programmatic efforts will be directed toward ensuring that all affected procedures are changed which are related to lessons learned or plant changes. As part of the Restart Operator Training Program, f the operations personnel will receive training on plant B-33
modifications, equipment changes, procedure revisions, and casualty control philosophy changes. FINDING - B.22: Operating Crew Minimum Required Staffing
- 22. Additional staffing above that required by plant Technical Specifications and other SMUD regulatory commitments allowed operators to perform certain tasks simultaneously. With staffing at the minimum required level, the actions performed would have had to be performed sequentially, would have taken longer, and could have exacerbated the overcooling transient.
DISTRICT RESPONSE The actions required to control the plant following the loss of ICS power event on December 26, 1985, could have been performed by the minimum required staff. The overcooling could have been terminated at any point by simply performing the E0P step to " trip the ap,ropriate pumps." It was a conscious decision by the operator to not perform that step. That decision process has been resolved by subsequent management policy, procedure changes, and operator training, as addressed in response to Finding 5. Furthermore, the District has implemented modifications and procedural changes which would decrease the demands upon the operators by providing controls in the control room which are powered independently of the ICS. This would eliminate the need to dispatch operators out into the plant. FINDING - B.23: Role of STA
- 23. Neither the operators nor the Shift Technical Advisor (STA) could identify an instance of when the STA provided engineering expertise during the incident. However, the operators found the STA valuable as an extra person on shift to help out during the incident.
DISTRICT RESPONSE Section 4.5 of NUREG-1195 indicates that the STA participated in the decision not to trip the Auxiliary Feedwater Pumps. That statement suggests that the STA did provide engineering input to the operators' decision-making process. However, the District has reaffirmed the role of the STA in the decision-making process to ensure that the STA , functions in an independent overview role and that the operators have *J engineering expertise available when needed. The STA provides 5 valuable support to the operators and the District intends to continue to support their utilization in this role. They are vital . members of the operating crew and can significantly enhance nuclear safety. The operator-on-watch standing principles includes training on the role and function of the STA. O B-34
FINDING - B.24: Application of Systematic Troubleshooting y
- 24. It appeared to the Team that SMUD personnel found the process of J troubleshooting in a highly controlled, systematic, and well-documented manner, as proposed by the Team, to be quite different from their usual maintenance practices. This difference contributed to the difficulty that the Team experienced in reviewing the troubleshooting program.
DISTRICT RESPONSE Immediately following a reactor trip on October 2, 1985, the District instituted a systematic program for analyzing the event and resolving root causes. The program was based on NUREG-il54 Appendix B, which described the Davis-Besse systematic troubleshooting program. The District's Transient Analysis Program was again implemented following a trip on December 5, 1985. Following the December 26, 1985 event, the District again implemented the systematic Transient Analysis Program for troubleshooting. The project was in full effect when the IIT arrived on site several days later. The IIT performed a line-by-line cocparison of the District's program with NUREG-Il54 Appendix B without considering procedural and organizational differences between the two organizations. As a result, the District's program was twice revised to incorporate IIT wording which, in th; District's opinion, did not constitute substantive changes affecting the outcome of the troubleshooting or the effectiveness of the program. The District had a highly controlled, systematic, and well documented troubleshooting program in place prior to, and following, the December 26, 1985 event. FINDING - B.25: IIT Requests for Information
- 25. of the December 26, 1985 incident, Throughout SMUD personnel the hadTeam's considerarevi@ble difficulty providing information in the detail that the Teamtrequested. Thus, SMUD personnel repeatedly summarized data," analyses, and plans without including the actual data and analyses. As a result, the Team had to request the detailed ' underlying data and analyses, which subsequently were provided. This iterative process delayed the Team's on-site investigation.
DISTRICT RESPONSE The Of. strict's investigations following the December 26, 1985 event were broad in scope and exceeded that identified by the IIT. As the IIT increased its knowledge of the plant and the event, they expanded their areas of interest. Often the District already had an investigation under way in the area of question. (This may have led the IIT to perceive that the District was not providing information O in the detail requested.) When requested, the detailed information Q was provided as stated in Finding 10 above. B-35 i l l
The District did have difficulty in anticipating the areas where the IIT would desire detailed information. Significantly, the detailed information was available when requested, indicating that the District had independently implemented an effective troubleshooting program. FINDING - B.26: Applicability of Generic PTS Analysis to Rancho Seco
- 26. In June 1983, the B&W Owner's Group reported (BAW-1791) the results of an analysis which predicted an overcooling transient caused by a loss of ICS power could occur at B&W-designed reactors with a high probability (about 4x10-2 per reactor year). If this probability were applicable to all eight B&W-designed operating reactors, such a transient could occur at some B&W-designed plants approximately every three years. Thus, it would appear that this analysis predicts that events comparable to the December 26, 1985 incident would occur approximately once every third year even if the EFIC system were installed at all B&W-designed plants. In addition, the report notes that one B&W-designed plant has a combination of components that cause the transient frequencies to be even higher. The Team deduced that the plant was Rancho Seco.
Finally, the generic B&W PTS analysis (BAW-1791) is not directly applicable to Rancho Seco because it assumes that the EFIC system is installed. DISTRICT RESPONSE The B&W Owners Group has reviewed BAH 1791 since the December 26, 1985 event. Overall, the report was found to still be valid. During the short term (prior to EFIC installation), the report underpredicts the frequency of occurrences for some overcooling events. The report becomes more representative with the installation of EFIC. With the PTS rulemaking in December 1985, the NRC established screening , criteria for PTS. The District's response to the screening criteria l has been submitted, and will be incorporated into the USAR in the 1986 update scheduled for July 1987 submittal. l l l l I O 1 B-36 l
O APPENDIX D ACTION PLAN COMMITMENT
SUMMARY
0 1
APPENDIX D O ACTION PLAN COMMITMENT
SUMMARY
The following items are summarized from the Rancho Seco Action Plan for Performance Improvement. The purpose of this summary is to provide a cross reference tool to assist in the closure of all Action Plan items.
- This summary is formatted to provide a section overview of commitments (when necessary), the Action Plan paragraph number and a description of the commitment.
Those individuals submitting restart recommendations per the QCI-12 procedure shall use this summary to cross reference their recommendation to the Action Plan. . NC - no change NA = not applicable 1 0 4 O D-1
1.4 SYSTEM REVIEH AND TEST PROGRAM Latest Amendment Amendment 1 NA 1. 4. 2.1 Main Feedwater System functional evaluation including reliability assessment will be performed. (PRIORITY 2) NA 1.4.2.2 Auxiliary Feedwater System functional evaluation including rollability assessment will be performed. (PRIORITY 2) NA 1.4.2.3 ICS/NNI functional evaluation including reliability assessment will be performed. (PRIORITY 2) NA 1.4.2.4 Pressure Control Functions of the Main Steam System functional evaluation including reliability assessment will be performed. (PRIORITY 2) NA 1.4.2.5 Instrument #1r System functional evaluation including reliability assessment will be performed. (PRIORITY 2) O l l O D-2 ( ----
4.A SYSTEMATIC ASSESSMENT PROGRAM Latest Amendment Amendment 1 NC 4.A.a. Prior to restart, systematic assessment programs will be established as a part of the administrative process of Rancho Seco. (PRIORITY 1) O O D-3
4A.1 PRECURSOR REVIEW PROGRAM Latest Amendment Amendment 1 NC 4A .1. 2.1. a . All Transient Assessment Program (TAP) Category C transients will be evaluated and investigated for their applicability and impact on Rancho Seco. (PRIORITY 1) NC 4A .1. 2.1. b . All Category B TAP events will be reviewed to determine if any of the recommendations made are applicable to Rancho Seco and to determine whether because of plant differences, the transient could have been more severe at Rancho Seco. (PRIORITY 1) NC 4A.1.2.1.c. All recommendations for Category A TAP transient will be reviewed to datermine their appilcability to Raacho Seco. (PRIORITY 1) NC 4A.1.2.1.d. All Rancho Seco transients, starting from the Rancho Seco " light bulb" event in 1978, will be reviewed. (PRIORITY 1) NC 4A .1. 2.1. e . Review NUREG-0667 " Transient Response of B&W designed Reactors" for open commitments. NC 4A .1. 2. 2. a . Rancho Seco Licensee Event Reports and Occurrence Description Reports will be reviewed. (PRIORITY 1) NC 4A.1.2.2.b. Significant Operating Experience Reports (SOER) issued by the Institute of Nuclear Power Operations will be reviewed. (PRIORITY 1). NC 4A .1. 2. 2. c . Bulletins issued by the NRC Office of Inspection and Enforcement will be reviewed. (PRIORITY 1) NC 4A.1.2.2.d. Notices / Circulars issued by the NRC Office of
- Inspection and Enforcement will be reviewed.
(PRIORITY 1) NC 4A.1.2.2.e. Babcock & Hilcox Reports (Preliminary Safety Concerns, Site Instructions, and other relevant BAH reports) will be reviewed. (PRIORITY 1) O i 0-4
.___ _. ._ _ _ ._ _ -. _ _ _ . _ . _ _ _~_ ._.. _ _ . _ _ _ _ _ .
l 4A.2 DETERMINISTIC FAILURE CONSEQUENCE ANALYSIS Latest-Amendment Amendment 1 NC 4A.2.3.1 An evaluation of the effects of loss of Electrical Power will be performed. (PRIORITY 1) NC 4A.2.3.2 An evaluation of the effects of loss of Instrument Air will be performed. (PRIORITY 1) NC 4A.2.3.3 The loss of ICS and NNI power supplies will be evaluated to determine failure states and resultant actions or suggested modifications necessary to establish a known safe state with little or no operator action. (PRIORITY 1) NC 4A.2.3.4 IE Bulletin 79-27 (Loss of Non-Class lE Instrumentation and Control Power Systems Bus During Operation) will be evaluated using a
; Deterministic Failure Consequences Analysis 2
approach. (PRIORITY 1) O
- O D-5
4A.3 B&W OWNERS GROUP PROGRAMS - SAFETY AND PERFORMANCE IMPROVEMENT PROGRAM (SPIP) Latest Amendment Amendment 1 NC 4A.3.2 The District will fully participate in the Safety and Performance Improvement Program. O 1 O D-6
. 4A.4 PLANT INTERVIEWS l Latest Amendment Amendment 1 i NC 4A,4.2 A minimum number of interviews will be ! established to identify systems, components, or operational problems and concerns of which they are aware and provide recommendations on how to ,' resolve them. (PRIORITY 1) t l i I 4 4 8 4 I
\
F e 1 1 I i N 6 k i l i i i r i I t I i ; i ! D-7 I i i
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4
- 48. MANAGEMENT, OPERATIONS AND ADMINISTRATIVE PROCESS IMPROVEMENT 48.1 MANAGEMENT EFFECTIVENESS Latest Amendment Amendment 1 NC 48.1.1.1 Review current executive level management practices and attitudes to ensure that executive level management processes support the safe and reliable operations of Rancho Seco. (PRIORITY 1)
NA 48.1.2.1.a. Establish within the Board of Directors, guidelines and agreements by which the Board, as an entity, can more effectively set policy and direction. (PRIORITY 2) NC 48.1.2.1.b. Establish written performance measurement criteria, and a performance process, for the General Manager (GM). (PRIORITY 2) NC 48.1. 2.1. c . Clarify the Board / General Manager working i relationship in writing, including the reporting desired by the Board from the General Manager. (PRIORITY 2) NC 48.1.2.2.a. Assess current corporate-support interfaces with the Nuclear Organization and make recommendations to the AGM, Nuclear and the GM regarding improved management of interorganizational working relationships. (PRIORITY 2) NC 48.1.3.a. Develop and implement a Rancho Seco Business Plan for use by the Board of Directors. (PRIORITY 2) NC 48.1.3.b. Establish a comprehensive, cohesive and clearly understandable set of GM and AGM-Nuclear policies and practices which provide upper tier direction for similar efforts at the functional manager and supervisory levels. (PRIORITY 2) NC 48.1.3.c. Establish up-to-date functional organization charters and position descriptions which accurately reflect responsibilities authorities, and accountabilities for all organization functions and job classifications. (PRIORITY 2) NC 48.1.3.d. Upgrade management programs and practices in the areas of functional planning, decision making, problem solving and interdepartmental collaboration. (PRIORITY 2) D-8
Latest m Amendment Amendment 1 NC 48.1.3.e. Establish appropriate management monitoring and control systems to ensure that all levels of department management are kept informed on important department performance trends or problem areas on a timely basis. At the same time, ensure that excessively burdensome administrative control systems are not perpetuated or introduced. (PRIORITY 2) NC 48.1.3.f. Develop an employee communications program originating from the office of the AGM-Nuclear to ensure that all department employees are kept informed of District concerns, departmental priorities and performance progress on a timely basis and encouraged to feel that they are an important part of the Rancho Seco team. (PRIORITY 2) NC 48.1.3.g. Develop a program for improving communications skills of Nuclear Department managers in presentations to the Board of Directors, the public, and staff. (PRIORITY 2) O NC 48.1.3.h. Establish a department Human Resource Management Q program which includes: (PRIORITY 2) NC 48.1.3.h. 1) identification of priority management development / training needs and the appropriate means for addressing each; NC 48.1.3.h. 2) identification of departmental priorities in terms of current vacancies and/or pipeline concerns; NC 48.1.3.h. 3) engage more department management collaboration with the District's Human Resources organization in the recruitment / selection process. NC 48.1.3.i. Improve Department media and community relations , by establishing a more proactive media / community l outreach program. (PRIORITY 2) NC 48.1.3.J. Improve Nuclear Department interfaces with all other Departments in the District by instituting additional interdepartmental communication and problem-solving processes on a regular basis. (PRIORITY 2) 48.1.3.k. Develop a Rancho Seco Facilities Master Plan. O NC , (PRIORITY 2) D-9 i
48.2 QUALITY AND QUALITY ASSURANCE Latest Amendment Amendment 1 NC 48.2.1.1 Reorganize the Quality function at Rancho Seco i to enhanca the Site Quality Assurance l Department, providing increased focuses in the following areas: (PRIORITY 1) 1 48.2.1.1.a. Quality Engineering l
- 48. 2.1.1. b . Quality Control 48.2.1.1.c. Surveillance !
- 48. 2.1.1. d . Vendor Qualification and l Source !
Inspection
- 48. 2.1.1. e . Nuclear Program Audits NC 48.2.1.2 Develop and implement the procedures and processes necessary to independently verify the effective closure of the actions identified for the Action Plan. The QTS tracking system has been developed to aid in completing this task.
(PRIORITY 1) NC 48.2.1.3 Institute interim measures to strengthen the materials control at the Rancho Seco site. This action will provide additional assurance that materials being installed are properly documented and in compliance with the applicable codes and standards. (PRIORITY 1) NC 48.2.1.4 Institute interim measures to enhance the integration of QC planning with maintenance and construction instructions and activities. (PRIORITY 1) NC 48.2.1.5 Increase the Site QA Department staff to assure the added demands of the Action Plan and changes in responsibilities can be effectively implemented. (PRIORITY 1) NA 48.2.2.1 Update and modify the Quality Program policies and procedures to enhance the effectiveness of the Quality Program, particularly those dealing with material control, engineering, quality surveillance, and maintenance. (PRIORITY 2) NA 48.2.2.2 Identify and develop enhancements to the QA program to address any programmatic and management areas that have identified deficiencies from a quality perspective. (PRIORITY 2) D-10
1 i Latest. Amendment Amendment 1 NA 48.2.3.1 Develop and implement the necessary policies and . procedures to establish a more proactive quality program, which will also improve the j effectiveness of audits. (PRIORITY 3) l I , l , I ( 1 I I \ 0-11
48.3 TRAINING Latest Amendment Amendment 1 NA 48.3.1.1.a. Continue the upgrade of Non-Licensed Operator Training to maintain INPO Accreditation. (PRIORITY 2) NA 48.3.1.1.b. Initiate the process of achieving INPO Accreditation for the maintenance training area. (PRIORITY 2) NA 48.3.1.1.c. Develop the plan for installation of a computerized Training Information Management System. (PRIORITY 2) NA 48.3.1.1.d. Develop plans for centralized and secure storage of training records. (PRIORITY 2) NA 48.3.1.1.e. Develop or purchase a Rancho Seco Simulator Baseline Data Information and Tracking System consistent with the Simulator now being purchased. (PRIORITY 2) NA 48.3.1.1.f. Incorporate short term training items (lessons learned) into permanent training materials. (PRIORITY 2) NA 48.3.1.2.a. Complete the purchase and installation of a plant specific simulator. (PRIORITY 3) NA 48.3.1.2.b. Complete the staffing of the Training Department with 3 MUD employees. (PRIORITY 3) NA 48.3.1.2.c. Complete and maintain INPO accreditation for the remainder of the Training Programs. (PRIORITY 3) 48.3.1.1 48.3.2.1.a. Train licensed operators on Emergency Operating 48.3.2.12 Procedures, including the changes resulting from 48.3.1.22 the December 26, 1985 event, those revisions 48.3.1.28 resulting from the E0P/ATOG review, and all 48.3.1.26 completed recent design modifications. 48.3.1.27 (PRIORITY 1) 48.3.1.2 48.3.2.1.b. Train licensed operators on the loss of ICS/NNI, including those procedures added or revised as a result of the December 26, 1985 event and related design modifications. (PRIORITY 1) 48.3.1.3 48.3.2.1.c. Train operators on valve applications and operation including OJT on local and/or manual operators. This includes limits and precautions such as use of valve wrenches. Incorporate into requalification training. (PRIORITY 1) D-12 e- ~r
Latest s Amendment Amendment 1 I'v) 48.3.1.5 48.3.2.1.d. Train operators on the specific lessons learned from the December 26, 1985 event. These include 48.3.1.6 48.3.1.13 such items as the makeup pump failure, 48.3.1.16 overfilling the makeup tank, cooldown rates, 48.3.1.17 reactor vessel head bubble information, and the 48.3.1.18 functioning of valve actuator controls. (PRIORITY 1) 48.3.1.4 48.3.2.1.e. Train operators on watch standin; principles, 48.3.1.14 including command and control training for shift supervisors, role and function of STA, equipment monitoring. (PRIORITY 1) 48.3.1.15 48.3.2.1.f. Retrain operators on health physics requirements associated with their job responsibilities. (PRIORITY 1) 48.3.2.1.g. Train operators on their job related functions associated with startup testing. (PRIORITY 1) 48.3.1.15 48.3.3.1.a. Train Health Physics Technicians and Operators on the procedure (s) for entry into areas of unknown radiological conditions. (PRIORITY 1) O 48.3.3.1.b. Train Health Physics Technicians and Operators h on proper response to radiological emergencies. (PRIORITY 1: 48.3.1.29 48.3.3.1.c. Train Health Physics Technicians and Operators on evaluation of radiological effluent discharges,
- i (PRIORITY 1) 48.3.4.1.a. Train assigned maintenance personnel on the maintenance of the Interim Data Acquisition and Display System (IDADS). (PRIORITY 1) l 48.3.4.1.b. Train operators on the operational use of IDADS.
- (PRIORITY 1) l 48.3.4.1.c. Update Emergency Preparedness Training l
Instructor Guides, Student Guides, and visual aids to support the October 1986 Drill. (PRIORITY 1) 48.3.1.24 48.3.4.1.d. Train assigned personnel on the revised Emergency Plan Procedures. (PRIORITY 1) 48.3.4.1.e. Provide management guidance to the operating Q crews (through training) on prioritizing multi-casualty events. (PRIORITY 1) Q D-13
Latest Amendment Amendment 1 NA 48.3.4.2.a. Develop and implement an improved process for continuing Emergency Response Organization training. (PRIORITY 3) O O. 0-14
a 48.4 OPERATIONS AND OPERATING PROCEDURES
'),- '
Latest
; Amendment Amendment 1 NC 48.4.1.1 Issue a procedure defining the policy for j procedural compilance and procedural guidance.
j This procedure will provide a direction on what i constitutes " procedural compilance" and j " procedural guidance." l (PRIORITY 1) i NC 48.4.1.2 Correct specific procedural deficiencies
! identified during the review of the December 26, i 1985 transient. ! (PRIORITY 1) 4C.28.27 48.4.1.3 Review and upgrade the CR/TSC HVAC operating j procedures. (PRIORITY 1) i j 48.4.1.3 48.4.1.4 Verify technical correctness of E0P changes made ; since May 1985. (PRIORITY 1) f 48.4.1.4 48.4.1.5 Compare E0Ps to ATOG Technical Basis, j Incorporate identified improvements into E0Ps. ; (PRIORITY 1)
Make the necessary modifications to the design f'3 48.12.1.2 48.4.1.6 change process to assure that design changes are
- Incorporated into all operating procedures in a timely manner.
(PRIORITY 1) 48.4.1.5 48.4.1.7 Assure operating procedures address the
! recommended topics of Regulatory Guide 1.33 Sections listed below. Implement procedures which may be required.
(PRIORITY 1) 48.4.1.7.a. Section 3 Procedures for Startup, Operation, and Shutdown of Safety Related Power System.
- 48. 4.1. 7. b . Section 6 Procedures for Combating Emergencies and Other Significant Events.
48.4.1.6 48. 4.1. 8. a Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Air Ejector / Gland Seal System. (PRIORITY 1) v l 0-15
Latest Amendment Amendment 1 48.4.1.6 48.4.1.8.b. Perform a valve wi.lkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Auxiliary Feedwater System. (PRIORITY 1) 48.4.1.6 48.4.1.8.c. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Auxiliary Steam System. (PRIORITY 1) 48.4.1.6 48.4.1.8.d. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Component Cooling Water System. (PRIORITY 1) 48.4.1.6 48.4.1.8.e. Perform a valve walkdown to verify the conststency of as-bullt conditions, P&I0s, procedural lineups, and component identification l for the Instrument Air System. (PRIORITY 1) 48.4.1.6 48.4.1.8.f. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Main Circulating Water System. (PRIORITY 1) 48.4.1.6 48.4.1.8.g. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Main Condensate System. (PRIORITY 1) 48.4.1.6 48.4.1.8.h. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Main Feedwater System. (PRIORITY 1) 48.4.1.6 48.4.1.8.i. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Nitrogen Gas System. (PRIORITY 1) 48.4.1.6 48.4.1.8.J. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Plant Cooling Water System. (PRIORITY 1) 48.4.1.6 48.4.1.8.k. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Service Water System. (PRIORITY 1) D-16
, Latest Amendment Amendment 1 48.4.1.6 48.4.1.8.l. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Turbine Electro Hydraulic Control System. (PRIORITY 1) 48.4.1.6 48.4.1.8.m. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Turbine Lube Oil System. (PRIORITY 1) 48.4.1.9 Establish an Administrative Procedure to ensure that " Systematic Troubleshooting" is accomplished on requisite future events.
Guidance for this program is provided in memo GAC 85-1001, Rev. 2. (PRIORITY 1) NA 48.4.2.1 Develop a revised Nuclear Operations organization and begin staffing at the management level. (PRIORITY 2) NA 48.4.2.2 Develop a staffing plan and schedule to meet the needs of the revised Nuclear Operations organization. This will include the needs for I licensed operators identified as rotational / transfer assignments. (PRIORITY 2) l 4 l l l IO 0-17
48.5 MAINTENANCE PROGRAMS AND PROCEDURES Latest Amendment Amendment 1 48.5.1.7 48.5.1.1 Inventory Calibrated Test Equipment (CTE) and calibrate and/or control use to prevent use of uncalibrated CTE. (PRIORITY 1) 48.5.1.8 48.5.1.2 Identify and assure current calibration of all in-plant instrumentation used in the performance of surveillance testing. (PRIORITY 1) 48.5.1.1 48.5.1.3 Rework the makeup pump and return to service. (PRIORITY 1) 48.5.1.11 48.5.1.4 Complete the in-orogress battery replacements (A, B, C, D, E, F). (PRIORITY 1) 48.5.1.2 48.5.1.5 Perform refueling interval surveillance of snubbers. (PRIORITY 1) 48.5.1.9 48.5.1.6 Complete rework of terminations in the Bailey l Cabinets in the Control Room (NNI/SFAS/RPS/ICS). (PRIORITY 1) 48.5.1.3 48.5.1.7 Perform biennial Diesel Generator Inspection and replace turbochargers. (PRIORITY 1) 48.5.1.4 48.5.1.8 Define the critical items to be included in the PM program. (This is to be an accelerated portion of the planned PM Program Upgrade.) As a minimum, this will include the Manual Limitorque Operated Valves (105), the Manual Non-Limitorque Operated Valves (135), other Manual Valves important to process flow control in Class I and steam generator heat removal applications (143), plant instrumentation required for surveillances, safety related HVAC and the Control Room normal HVAC system. (PRIORITY 1) 48.5.1.6 48.5.1.9 Complete Preventive Maintenance (PMs) on selected manual valves identified in 48.5.1.8 above. (PRIORITY 1) NA 48.5.2.1 Develop a departmental procedure hierarchy and writer's guide for Maintenance Procedures. (PRIORITY 2) NA 48.5.2.2 Identify and prioritize maintenance procedures for generation and/or revision. (PRIORITY 2) l D-18 j i 1
Latest g Amendment Amendment ', NA 48.5.2.3 Achieve authorized staffing levels within the PM organizations and activities. (PRIORITY 2) . NA 48.5.2.4 'l Develop and/ororevise'the required programmatic
" procedures f6r the PM program to: assign responsibilities, author,ity and accountabilities for the program; establish criteria and define
- the scope of the program; and define the interface with other work control processes.
(PRIORITY 2) ,.
- zu ,
.NA 48.5.2.5 Review existing PM tasks and frequency for critica1' equipment. Revise and augment as required by programmatic selection criteria.
,,(PRIORITY.*2) ..
- r. p Perform Laboratory Failure, Analysis of the ICS NA 48.5.2.6 /*
- 51/S2 switches and ICS Power Supply Monitor
- J' which were 1.n place on December 26, 1985.
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48.6 HEALTH PHYSICS AND RADIOLOGICAL CONTROLS Latest i Amendment Amendment 1 48.6.1.1 48.6.1.1 Relieve operators of special HP duties. (PRIORITY 1) 48.6.1.2 48.6.1.2 Prepare a procedure, for Health Physics Technicians to use for entry into unknown radiological conditions. (PRIORITY 1) 48.6.1.3 48.6.1.3 Revise setpoints for plant gaseous effluent monitors to ensure unambiguous indications. (PRIORITY 1) 48.6.1.4 48.6.1.4 Issue a Radiological Event Directions Manual to provide more guidance for abnormal situations. (PRIORITY 1) 48.6.1.5 48.6.1.5 Issue new manuals to separate event and instrument procedures from the Radiation Control Manual. (PRIORITY 1) 48.6.1.6 Organize and maintain the Health Physics Support and Environmental Programs group. 48.6.1.7 Issue revised ALARA Manual. 48.6.1.8 Develop technical support and auditing management processes and procedures to interface HPSEP and Radiation Protection groups. 48.6.1.9 Respond to NRC Violation Enforcement 86-06-05: Evaluate actions with implementation of AP.305 and H2 PSA-7. I O D-20
l 48.7 10CFR50 APPENDIX I DISCHARGE GUIDELINES ] l O' Latest Amendment Amendment 1 48.7.1.5 48.7.1.1 Evaluate the current radioactive waste analysis methods and sensitivity relative to their ability to support operation needs and to provide confidence that discharges and the cumulative impact of discharges will satisfy the objectives of the environmental discharge requirements. (PRIORITY 1) 48.7.1.5 48.7.1.2 Develop and implement the changes in 48.7.1.11 Radiochemistry methods and controls necessary to provide confidence that discharges and the cumulative impact of discharges will satisfy the objectives of the environmental discharge requirements. (PRIORITY 1) 48.7.1.2 48.7.1.3 Review and revise the off-site discharge 48.7.1.11 calculation manual to incorporate changes in 48.7.1.14 Radiochemistry methods and controls necessary 48.7.1.17 per 4B.7.1.2. (PRIORITY 1) NA 48.7.2.1- Evaluate the design of plant systems with the intent to improve the ability to operate within Appendix I Guidelines when operating with O' - primary to secondary leakage. Implement plant improvements as appropriate. (PRIORITY 2) D-21
48.8 EMERGENCY PREPAREDNESS Latest Amendment Amendment 1 48.8.1.1 Meet with NRC (Region V) to review / critique Emergency Action Plan. (PRIORITY 1) 48.8.1.1 48.8.1.2 Update Emergency Plan and Implementing Procedures to address December 26 event lessons Learned. (PRIORITY 1) 48.8.1.3 Establish independent meteorological assessment capability. (PRIORITY 1) 48.8.1.4 Integrate EP commitments into commitment tracking program. (PRIORITY 1) 48.8.1.2 48.8.1.5 Evaluate notification / communication system and implement upgrades related to December 26 event Lessons Learned. (PRIORITY 1) 48.8.1.6 Simplify Control Room dose calculation procedure. (PRIORITY 1) 48.8.1.7 Implement new PASS procedures including core damage assessment. (PRIORITY 1) 48.8.1.8 Initiate " mini-drills" program for emergency preparedness. (PRIORITY 1) 48.8.1.3 4B.8.1.9 Coordination and support of the Training Group per 48.3 including: (PRIORITY 1) . 48. 8.1. 9. a . Emergency Response Organization ! (ERO) identification
- 48. 8.1. 9. b . Plan / procedure update information 4B . 8.1. 9. c . Facilities identification
- 48. 8.1. 9. d . Instruction materials upgrade 48.8.1.9.e. Scheduling l 4B . 8.1. 9. f . Tracking and documentation 48.8.1.9.g. Data base management
- 48. 8.1. 9. h . Management support of ERO participation NA 48.8.2.1 Establish separate onsite and corporate plans for emergency preparedness. (PRIORITY 2)
NA 43.8.2.2 Consolidate / cross index Emergency Preparedness (EP) and Central files. (PRIORITY 2) NA 4B.8.2.3 Complete multi-parametric data base including EP records and schedules. (PRIORITY 2) 0-22
Latest Amendment Amendment 1 NA 48.8.2.4 Provide positional analyses for ERO versus District staff. (PRIORITY 2) NA 48.8.2.5~ Define / implement public education program enhancements for emergency response. (PRIORITY 2) NA 48.8.2.6 Integrate Emergency Plan Implementing Procedures and plant operating procedures. (PRIORITY 2) NA 48.8.2.7 Complete installation of notification / communication system including verification, training and drills program. (PRIORITY 2) NA 48.8.2.8 Redefine Emergency Preparedness maintenance program. (PRIORITY 2) 1 O O 0-23
48.9 HUMAN FACTORS Latest Amendment Amendment 1 48.9.1.1 Implement modifications and procedure changes resulting from the post trip Human Factors review including: (PRIORITY 1) 4B.3.1.3 4B.9.1.1.a. Provide operator training associated with AFW valves FV-20527 and FV-20528. i 40.33.8 4B.9.1.1.b. Provide accurate local hand jack position indication for AFH valves FV-20527 and FV 20528. 48.9.1.1.c. Improve interface between Security and Control Room I personnel. l 48.9.1 48.9.1.1.d. Install long cord on red phone. 40.1.13 48.9.1.1.e. Relabel ICS power supply breakers Sl/S2. NA 48.9.2.1 Modify Control Room access doors such that one door is used for exit and one for entrance. (PRIORITY 2) NA 48.9.2.2 Modify red phone power supply to eliminate i spurious ringing following power supply transfers. (PRIORITY 2) NA 48.9.2.3 Assess capability to accelerate implementation , of Control Room Design Review (CRDR) modifications (MOD 142) currently scheduled for , Cycle 8 and Cycle 9 refueling outages. If possible accelerate implementation. (PRIORITY 2) NA 48.9.3.1 Participate in the INPO Human Performance l Evaluation program. (PRIORITY 3) NA 48.9.3.2 Implement CROR modifications (MOD 142). These modifications were identified in the District's submittal to the NRC in December, 1985, which documented the results of the Rancho Seco Control Room design review. (PRIORITY 3) NA 48.9.3.3 Review modifications incorporated following 12/26/85 event for impact and/or compliance with CROR criteria / program. (PRIORITY 3) 0-24
48.10 MANAGEMENT INFORMATION SYSTEM b Latest Amendment Amendment 1 48.10.1.1 Review implementation of NRC Generic Letter 83-28 commitments and develop plan for program enhancement. (PRIORITY 1) 48.10.1.1 48.10.1.2 Provide site information system support for implementation and records management activities needed for restart. (PRIORITY 1) NA 48.10.2.1 Prepare a Nuclear Information Systems General Design, identifying and prioritizing improvement projects for implementation within three years. (PRIORITY 2) NA 48.10.2.2 Implement Vendor Data Program enhancements identified to achieve the program objectives. (PRIORITY 2) NA 48.10.3.1.a. Complete Nuclear Information Management System (NIMS) evaluation. (PRIORITY 3) NA 48.10.3.1. b . Establish on-site facilities and organization to support NIMS hardware / software. (PRIORITY 3) NA 48.10.3.1.c. Implement NIMS program / data. (PRIORITY 3) O D-25
48,11 COMMITMENT MANAGEMENT Latest Amendment Amendment 1 J NC 48.11.1.1 Revise commitment management procedure to include tracking and compliance features. ' (PRIORITY 1) NC 48.11.1.2 Install a new commitment tracking system per , 48.11.1.1. (PRIORITY 1) 48.11.1.3 Develop system / user documentation for the new comm!tment tracking system. (PRIORITY 1) NC 48.11.1.4 Verify the commitment tracking system database with respect to current known commitments. (PRIORITY 1) NC 48.11.1.5 Verify all commitments required prior to restart are complete. (PRIORITY 1) 48.11.2.1 Verify the comprehensiveness of the commitment tracking system database. (PRIORITY 2) 48.11.2.2 Each Nuclear Department will establish specific milestones for reducing its backlog of open commitments. (PRIORITY 2) 48.11.3.1 Integrate the commitment tracking system with the Nuclear Information Management System. (PRIORITY 3) 0 0-26
48.12 CONFIGURATION MANAGEMENT
/ Latest Amendment Amendment 1 NC 48.12.1.1 Verify that Control Room drawings are current, in accordance with existing procedures.
(PRIORITY 1) 48.12.1.2 48.12.1.2 Provide Nuclear Engineering support to plant 48.12.1.3 operations to address and expedite configuration management issues. (PRIORITY 1) NA 48.12.2.1 Review and evaluate all temporary modifications and close out all existing abnormal tags that need to be converted to permanent plant modifications. (PRIORITY 2) NA 48.12.2.2 Reduce the backlog of DCNs. (PRIORITY 2) NA 48.12.2.3 Develop the System Design Basis documents for NEP manuals on BOP related systems. (PRIORITY 2) NA 48.12.3.1 Establish management direction for a Configuration Management program for Rancho Seco consisting of: (PRIORITY 3) 48.12. 3.1. a . Policy 48.12.3.1. b . Specifications 48.12. 3.1. c . Computer hardware and software 48.12. 3.1. d . Implementing procedures 48.12. 3.1. e . Training NA 48.12.3.2 Establish or upgrade existing equipment and supporting documentation identification systems needed for total plant configuration control. (PRIORITY 3) NA 48.12.3.3 Upgrade change control packages that control modification from change request through close out. This includes identification of all affected documentation such as procedures, training plans, and simulator upgrades. (PRIORITY 3) NA 48.12.3.4 Reorganize drafting into a design / drafting organization. (PRIORITY 3) NA 48.12.3.5 Train Nuclear Engineers to utilize the designers
- from the design / drafting organization identified in 48.12.3.4 to reduce the engineering work p
load. (PRIORITY 3) D-27
Latest Amendment Amendment 1 NA 48.12.3.6 Develop new, or simplify existing procedures to clearly define the review process for drawings. (PRIORITY 3) NA 48.12.3.7 Conduct a cost / schedule review to determine whether a CAD system can be justified for Rancho Seco. (PRIORITY 3) NA 48.12.3.8 Develop or upgrade existing systems to provide verification that configuration documentation reflects the true hardware configuration. (PRIORITY 3) NA 48.12.3.9 Develop or upgrade existing systems to provide the status of all documentation and equipment in a timely and accurate manner. (PRIORITY 3) NA 48.12.3.10 Develop a work package system for all facility changes. (PRIORITY 3) O O D-28
48.13 MATERIALS MANAGEMENT Latest Amendment Amendment 1 NA 48.13.1.1 Conduct a review of the current Materials Management program. (PRIORITY 2) i ; NA 48.13.1.2 Develop and implement an action plan to improve 4 the performance of the Materials Management 1 program. (PRIORITY 2) a a I i i ) i 5 a f r i l l 0-29
4C PLANT MODIFICATIONS AND MAINTENANCE IMPROVEMENTS 4C.1 INTEGRATED CONTROL SYSTEM (ICS) AND INTERFACING SYSTEMS { Latest Amendment Amendment 1 4C.1.16 4C.1.a.1.a.1 Improve reliability of ICS power supplies. (PRIORITY 1) 4C.1.a.1.a.1.a. Provide dedicated AC Inverters. 4C .1. a .1. a .1. b . Provide larger DC capacity. 4C.1.a.1.a.2 Implement Upgrades (PRIORITY 1) 4C.1.a.1.a.2.a. Reduce MFHP runback rate to 25%/ min. 4C .1. a .1. a . 2. b . Substitute for RC flow the RCP status, i.e., number of RCP's. 4C.1.a.1.a.2.c. Remove SU FW flow correction to the MFH flow signal. 4C.1.a.1.a.2.d. Replace modules with improved Bailey revised modules. 4C.1.a.1.a.2.e. Remove FW temp correction from total FW demand signal. 4C.1.a.1.a.2.f. Revise RCP Runback rate to 25%/ min. 4C.1.a.1.a.2.g. Revise asymetric CRD Runback rate to 3%/ min. 4C.1.8 4C.1.a.1.a.2.h. Improve post maintenance testing and surveillance. 4C.1.13 4C.1.a.1.a.2.1. Replace all S1 and S2 switches with new switches. 4C.1.12 4C.1.a.1.a.2.J. Rewire TCS Power Supply Monitor to delete the " Daisy chain." 4C.1.a.1.a.2.k. Test / Inspect all cabinet wirewraps and lugs. 4C.1.4 4C.1.a.1.a.3 Inspect electrical terminations within ICS. (PRIORITY 1) 4C.1.6 4C.1.a.1.b.1 Upgrade ICS annunciation to remove ambiguitles. 4C.1.23 (PRIORITY 1) 4C.1.a.1.b.2 Provide computer inputs for ABT (AC power supply) status. (PRIORITY 1) 4C.1.25 4C.1.a.1.b.3 Add open/close status lights for ADVs, TBV's, MFH valves. (PRIORITY 1) 4C.1.a.1.b.4 Tag instruments with ICS processed signals. (PRIORITY 1) D-30
Latest p Amendment Amendment 1 4C.1. a .1. b . 5 Provide first-out MFP trip monitor. (PRIORITY 1) 4C.1.a.1.b.6 Provide adequate parameter trending independent of ICS/NNI. (PRIORITY 1) 4C.1.13 4C.1.a.1.b.7 Provide clear labels for all Si and S2 switches. (PRIORITY 1) 4C.1.29 4C.1.a.1.b.8 Add all hot shutdown related parameters to SPDS, independent of ICS/NNI. (PRIORITY I) 4C.1.3 4C.1.a.1.c.1 Review procedures for adequacy in event of ICS failure. (PRIORITY 1) 4C.1.15 4C.1.a.1.c.2 Review / adjust minimum MFP speed. (PRIORITY 1) 4C.1.2 4C.1.a.1.c.3 Trip ICS power on loss of NNI-X. Y, or Z power. (PRIORITY 1) 4C.1.1 4C .1. a .1. c . 4 Trip MFP's on loss of ICS control. (PRIORITY 1) 4C.1.a.1.c.5 Provide independent control-grade backup AFH automatic level control. (PRIORITY 1) IAS 4C.1.a.1.c.6 Provide Class I bottled air supply for ADV's, AFH, MFH, and SUFW valves. (PRIORITY 1) 4C.1.7 4C.1.a.1.c.7 Modify Auxiliary Steam Reducing Station to fall at setpoint. (PRIORITY 1) 4C.1.3 4C.1.a.1.d.1 Review ICS restoration procedures for adequacy, correct as necessary. (PRIORITY 1) l 4C .1. a .1. d . 2 Evaluate (and correct as necessary) final control element position on ICS power restore. (PRIORITY 1) NA 4C.1.a.2.1 Develop and implement those design changes or enhancements that were not required to be implemented prior to restart, consistent with their assigned priority. (PRIORITY 2) NA 4C .1. a . 3.1 Actively participate in the B&W Owners Group efforts to upgrade or replace the Integrated l Control System. (PRIORITY 3) O ! 0-31
Latest Amendment Amendment 1 4C.1.b.1.1 Provide controls in the Control Room from which the operator can operate the TBV's, and which will cause these valves to remain closed on loss of ICS (DC) power. Note: ADV's will be controlled by EFIC. (PRIORITY 1) 4C.1.5 4C.1. c .1.1 Determine the contribution the ICS Power Supply Monitor (PSM) had in December 26, 1985 transient. (PRIORITY 1) 4C.1.c.1.2 Evaluate the potential improvement to ICS reliability if redundant PSM's are installed. (PRIORITY 1) 4C .1. c .1. 3 Determine the potential benefits to be obtained through the installation of independent ICS PSM's. (PRIORITY 1) 4C.1.c.1.4 Implement design improvements or document justification for not implementing ICS PSM modifications. (PRIORITY 1) 4C.1. d .1.1. a Provide a window for the status indication of ICS Trouble (fan failure, power supply failure). (PRIORITY 1) 4C.1. d .1.1. b Provide a window for the status indication of ICS failure (loss of DC buss). (PRIORITY 1) NA 4C.1.e.1.a Participate in B&W Owners Group efforts to perform a generic failure / consequence evaluation of the Model 820 ICS. This effort was initiated August 1, 1986. (PRIORITY 2) NA 4C.1.e.2.a Evaluate results and applicability of B&W Owners Group evaluation findings and recommendations to Rancho Seco. (PRIORITY 3) NA 4C.1.e.2.b Conduct supplemental evaluations as required to achieve Rancho Seco specific information. (PRIORITY 3) NA 4C.1.e.2.c Develop and implement applicable modifications to Rancho Seco ICS. (PRIORITY 3) 0 0-32
Latest Amendment Amendment 1 NA 4C.1.f.1.a Evaluate need for DC Bus battery backup based on reliability of power supplies, recent modifications, etc. (PRIORITY 3) NA 4C .1. f .1. b Evaluate Hand / Auto station backup power. (PRIORITY 3) NA 4C .1. f .1. c Participate with B&W Owners Group to enhance ICS and related power supplies. (PRIORITY 3) NA 4C .1. f .1. d Develop and implement ICS modifications identified as necessary during evaluations. (PRIORITY 3) l D-33 m--.~~-tw,,- em-ww- mme ~ m,,w,w,,,,
1 4C.2 NON-NUCLEAR INSTRUMENTATION (NNI) Latest Amendment Amendment 1 NA 4C.2.a.1.a.1 Improve reliability of NNI power supplies. (PRIORITY 2) NA 4C.2.a.1.a.2 Upgrade NNI Hodules to latest B&W models. (PRIORITY 2) NA 4C.2.a.1.a.3 Update NNI drawings to correct discrepancies identified in the deterministic failure analysis and the District's ongoing efforts to upgrade plant performance. (PRIORITY 2) 4C.1.24 40. 2. a .1. b .1 Upgrade NNI annunciation. (PRIORITY 2) NA 40.2.a.1.b.2 Provide computer inputs for NNI ABT status. (PRIORITY 2) NA 4C.2.a.1.b.3 Tag Control Room instruments with relevant power or signal channel identification. (PRIORITY 2) NA 4C.2.a.1.b.4 Provide adequate NNI parameter trending in IDADS (including recorder replacement). (PRIORITY 2) 4C.1.2 4C.2.a.1.c.1 Provide automatic trip of ICS power on loss of NNI-X, Y, or Z power. (PRIORITY 2) 4C.1.26 4C.2.a.1.c.2 Revise procedures for loss of NNI. (PRIORITY 2) 4C.1.32 4C.2.a.1.c.3 Review NNI restoration procedures for adequacy in event of NNI failure. (PRIORITY 2) NA 4C.2.a.2.1 Actively participate in the B&W Owners Group efforts to upgrade or replace the Non-Nuclear Instrumentation equipment. (PRIORITY 3) 4C.2.b.1.1 Evaluate the potential improvement to NNI reliability if redundant PSM's are installed. (PRIORITY 1) 4C.2.b.1.2 Determine the potential benefits to be obtained through the installation of independent NNI PSM's. (PRIORITY 1) 4C . 2. b .1. 3 Implement design improvements or documat justification for not implementing NNI PSM modification. (PRIORITY 1) O., 0-34
Latest Amendment Amendment 1 4C.1.30 4C.2.b.1.4 Inspect electrical terminations within NNI cabinets. (PRIORITY 1) 4C.2.c.1.1 Tag NNI-affected indicator / recorder. (PRIORITY 1) 4C.1.2 4C.2.c.1.2 Trip ICS power on loss of NNI-X, -Y, or -Z power. (PRIORITY 1) 4C.2.c.1.3 Make modifications to ICS to provide automatic control of ICS/NNI power. (PRIORITY 1) 4C.2.c.1.4 Provide separate annunciator windows to indicate: (PRIORITY 1) 4C . 2. c .1. 4. s Loss of NNI-X (DC) 4C . 2. c .1. 4. b Loss of NNI-Y (DC) 4C . 2. c .1. 4. c Loss of NNI-Z (DC, switching supply) 4C . 2. c .1. 4. d NNI trouble (fan failure, single power supply failure) O 0-35
4C.3 FEEDWATER AND STEAM SYSTEMS Latest Amendment Amendment 1 4C.3.1 4C.3.a.1.1 Install EFIC and provide a Control Room panel as an extension to the HISS pilel. (PRIORITY 1) 4C.3.a.1.2 Implement control grade modifications to close TBV's on loss of ICS power. (PRIORITY 1) 4C.25.15 4C . 3. b .1. a Develop and implement plant mcdifications to the auxiliary steam controls to assure valve control on loss of ICS power. (PRIORITY 1) 4C.3.c.1.1 Review operation of Main Feedwater Pumps and status of incorporating lessons learned from October 2, 1985 event. (PRIORITY 1) 4C.3.c.1.2 Retune the ICS to reduce Main Feedwater System contributions to reactor trips. (PRIORITY 1) 4C.3.c.1.3 Validate setpoints and proper initiation / interface with Auxiliary Feedwater System. (PRIORITY 1) 4C.3.c.2.1 The recommendations contained in the B&WOG Availability Committee Report, 47-1159449-00, "MFH Pump Trip Reduction Program Final Report" will be evaluated for applicability to Rancho Seco and implemented as appropriate. (PRIORITY 2) 4C.3.d.1.1 Halkdown the Main Steam Lines and verify that each service connection, greater than two inch diameter, is provided with capability to Isolate from the Control Room. (PRIORITY 1) 4C.3.d.1.2 Tag Control Room switches to clearly indicate valves associated with A- or B-Steam Generators (PRIORITY 1) 4C.3.d.1.3 Evaluate Main Steam Line Supports for effects of flooding. (PRIORITY 1) 4C.3.d.1.4 Increase IDADS sample frequency for Main Steam / Main Feedwater Parameters. (PRIORITY 1) O D-36
4C.4 EMERGENCY DIESEL GENERATOR RELIABILITY O Latest V Amendment Amendment 1 4C.4.1.1 Replace Emergency Diesel Generator Turbochargers. (PRIORITY 1) NA 4C . 4. 2.1 Evaluate performance history of the Bruce-GM Emergency Diesel Generators and develop recommendations for reliability improvement. (PRIORITY 2) NA 40.4.2.2 Determine and implement identified emergency diesel generator modifications. (PRIORITY 2) NA 4C.4.2.3 Enhance the emergency diesel generator preventive maintenance program. (PRIORITY 2) i l l l l l l l l D-37
4C.5 REACTOR COOLANT SYSTEM AND PRESSURIZER Latest Amendment Amendment 1 4C.22.18 4C.S.a.1.1 Issue ECNs for new pressurizer relief valve discharge piping supports and support modifications. (PRIORITY 1) 4C.22.18 4C .S . a .1. 2 Inspect, reanalyze, and redesign (as required) ring structure anchoring pressurizer relief valve supports to Pressurizer. (PRIORITY 1) 4C.22.18 4C.S . a .1. 3 Construct new pressurizer relief valve supports and modify existing supports and ring structure (if required). (PRIORITY 1) 40.22.18 4C .S . a .1. 4 Construct pressurizer support structure modifications and modifications to existing work platforms required to resist new pipe support loads. (PRIORITY 1) O O D-38
4C.6. ENHANCE THE POST ACCIDENT SAMPLING SYSTEM (PASS) OPERABILITY Latest [d Amendment Amendment 1 4C.41.2 4C.6.1.1 Complete the SCAS panel rebuild. (PRIORITY 1) 4C.41.2 4C . 6.1. 2 Complete associated peripheral PASS equipment upgrades. (PRIORITY 1) 4C.41.1 4C . 6.1. 3 Document the compensating equipment in the environmental lab. (PRIORITY 1) 4C.41.3 4C.6.1.4 Complete work required to solve H2 monitoring heat tracing problems. (PRIORITY 1) 4C,41,3 4C.6.1.5 Revise operating procedures and complete training on revised system and conduct system functional test. (PRIORITY 1) NA 4C.6.2.1 Replace Dionex program controller. (PRIORITY 2) NA 40.6.2.2 Install R-15044 Sample Dryers. (PRIORITY 2) NA 4C.6.3.1 Complete PASS decay heat valve replacement during the Cycle 8 outage. (PRIORITY 3) l O D-39
~
4C.7 ACTIONS TO ENHANCE CONTROL ROOM /TSC AND NSEB HVAC - OPERABILITY AND RELIABILITY Latest Amendment Amendment 1 4C.48.2 4C.7.1.1.a. Prepare and implement a detailed action plan / testing program to identify excessive Control Room noise source (s) and propose modifications. (PRIORITY 1) 4C.48.2 4C.7.1.1.b. Develop and implement those design and procedural changes needed to reduce noise levels to allowable limits. (PRIORITY 1) 4C.48.13 4C.7.2.1 Evaluate and implement design changes if 4C.48.14 necessary to improve balancing capabilities. (PRIORITY 2) NA 4C.7.2.2 Develop and implement the changes to install flow meters to facilitate surveillance testing of the Control Room /TSC HVAC filter units. (PRIORITY 2) NA 4C.7.2.3 Develop and implement the changes identified as necessary to facilitate maintenance of Control Room /TSC HVAC equipment (i.e., replace Air Handler Unit access doors, modify the lube
. manifold to condenser fans, etc.) (PRIORITY 2)
NA 4C.7.2.4.a. Develop and implement the changes necessary to add dampers through the TSC ceiling. (PRIORITY 2) 40.48.12 4C.7.2.4.b. Develop and implement the changes necessary to upgrade dampers in the wall between Control Room and TSC. (PRIORITY 2) 4C.48.8 4C.7.2.5.a. Investigate flow control through filter units 4C.48.23 and recommend improvements. (PRIORITY 2) 4C.48.37 4C.7.2.6.a. Investigate replacement of existing essential HVAC compressor motors. (PRIORITY 2) NA 4C.7.2.7.a. Develop improved methods to adjust NSEB essential air handler air flow. (PRIORITY 2) 4C.48.1 4C.7.2.8 If requirej, develop and implement design change to add a sample manifold to filter banks in each , of two u. tits. (PRIORITY 2) ! D-40 U
4C.8 INSTRUMENT AIR SYSTEM RELIABILITY Latest g- Amendment Amendment 1 4C.32.2 4C.8.1.1 Complete IAS system review to identify hardware modifications required to improve system reliability. (PRIORITY 1) 4C.32.4 4C.8.1.2 Replace leaking letdown filter valve operators. (PRIORITY 1) 4C.32.5 4C.8.1.3 Add diesel-driven air compressor. (PRIORITY 1) 4C.32.1 4C.8.1.4 Provide bottled air backup to critical valves. 4C.32.8 (PRIORITY 1) 4C.32.7 4C.32.11 4C.32.2 4C.8.1.5 Perform IAS walkdown to identify additional air leaks and any P&ID discrepancies. (PRIORITY 1) 4C.8.1.6 Develcp and initiate Priority 1 modifications identified during system review. (PRIORITY 1) 4C.8.2.1 Develop and implement Priority 2 and 3 modifications identified in IAS review. 'O (PRIORITY 2) l 4 0 0-41
4C.9 REACTOR BUILDING PURGE FLOW RATE MEASUREMENTS Latest Amendment Amendment 1 4C.9.1.1 Perform engineering evaluation of flow measuring system deficiencies and identify appropriate modifications. (PRIORITY 2) 4C . 9.1. 2 Install and test modifications identifled during engineering evaluation. (PRIORITY 2) O O 0-42
4C.10 FIRE PROTECTION SYSTEMS [ ') Latest y/ Amendment Amendment 1 4C.45.3 4C.10.a.1.1 Develop and implement modifications to provide for manual operator override of a trip of the Auxillary Building ventilation fans. (PRIORITY 2) 4C.10.a.1.2 Upgrade the control logic and schematic diagrams for the fire protection system. (PRIORITY 2) 4C.45.1 4C.10.a.1.3 Develop and implement appropriate upgrades to prevent spurious signals of fire alarm systems on power transients. (PRIORITY 2) 4C.45.8 4C.10.b.1.1 Develop and laplement fire alarm and HVAC panel upgrades to separate the control circuitry and equipment for the Train A and Train 8 Dampers. (PRIORITY 2) NA 4C.10.c.1.1 Identify vital areas of potential impact due to leakage through floors following actuation of fire protection systems. (PRIORITY 2) NA 4C.10.c.1.2 Evaluate effect of impact of potential leakage - on safe shutdown equipment. (PRIORITY 2) ( \- NA 4C.10.c.1.3 Inspect all vital electrical equipment areas for potential leakage paths. (PRIORITY 2) NA 4C.10.c.1.4 Evaluate the results of the leakage inspection and identify recommended corrective actions. (PRIORITY 2) NA 4C.10.c.1. 5 Develop and implement changes necessary to address recommended corrective actions from leakage inspection evaluation. (PRIORITY 2) NA 4C.10.c.1.6 Review and upgrade as necessary preventative maintenance procedures to maintain drain Ilnes clear of obstructions. (PRIORITY 2) 0 0-43
4C.11 MOTOR OPERATE 0 VALVES Latest Amendment Amendment 1 48.5.1.6 4C.11.1.1 Refurbish the 100 Safety Related MOV's. (PRIORITY 1) 4C.11.2.1 Refurbish the 63 Non-Safety Related MOV's. (PRIORITY 2) O O D-44
4C.12 CRITICAL PUMPS FAILURE ON LOSS OF SUCTION Latest O. Amendment Amendment 1 I 48.3.1.16 4C.12.1.1 Evaluate procedures and provide training to prevent recurrence of loss of suction pump failure. (PRIORITY 1) 4C.12.2.1 Engineering is to review design philosophy for suction valve interlocks and alarms on critical , pumps and identify appropriate modifications. ' (PRIORITY 2) i O O 0-45
4C.13 MAINTENANCE PROGRAMS AND ACTIONS Latest Amendment Amendment 1
- 48. 5.1.7 4C.13.1.1 Inventory Calibrated Test Equipment (CTE) and calibrate and/or control use to prevent use of uncalibrated CTE. (PRIORITY 1) 48.5.1.8 4C.13.1. 2 Assure current calibration of all in-plant instrumentation used in the performance of surveillance testing. (PRIORITY 1) 48.5.1.1 4C.13.1.3 Rework the makeup pump and return to service.
(PRIORITY 1) 48.5.1.11 4C.13.1.4 Complete the in-progress battery replacements (A, B, C, D, E, F). (PRIORITY 1) 4C.13.1.5 Perform refueling interval surveillance of snubbers. (PRIORITY 1) 48.5.1.9 4C.13.1.6 Complete rework of terminations in the Bailey Cabinets in the Control Room (NNI/SFAS/RPS/ICS). (PRIORITY 1) 48.5.1.3 4C.13.1.7 Perform blennial Olesel Generator Inspection and replace turbochargers. (PRIORITY 1) 48.5.1.4 4C.13.1.8 Define the critical items to be included in the PM program. (This is considered to be an accelerated portion of the planned PM Program Upgrade.) As a minimum, this will include the Manual Limitorque Operated Valves (105), the Manual Non-Limitorque Operated Valves (135), other Manual Valves important to process flow control in Class I and steam generator heat removal appilcations (143), plant instrumentation required for surveillances, safety related HVAC and the Control Room normal HVAC system. (PRIORITY 1) 48.5.1.5 4C.13.1.9 Complete Preventive Maintenance (PMs) on manual valves selected due to their functional position, e.g., Isolation of active equipment such as pumps, control valves, heat exchangers, ' cross-ties. (PRIORITY 1) 4C.13.1.10 Repair valves investigated as a part of December 26, 1985 event troubleshooting. Includes FV-20527, FV-20528, FHS-063, FHS-064. (PRIORITY. 1) 0-46
LL '/,4 Jf ,. t r
+
Y ,?
, Latest .
/ -
p A doent Amendment 1 m re ,
. NA' ; 4C.13.2.1 , Develop a departmental procedure hierarchy and
*v (. writer's guide for Maintenance Procedures.
i e 2 PRIORITY 2) s , / .- u, ,s , > NAS, 4C.13.2.z , ' Identify and prioritize maintenance' procedures p ' for generation and/or revision. (PRIORITY 2) NA' x 4C.13.2.3 Achieve authorized staffing levels within the i t j, maintenance organizations. (PRIORITY 2) NA ' 4C.13.2.4 Develop and/or revise the required programmatic [ y procedures for the PM program to: assign 3
~
/ responsibilities, authority and accountabilities
; for the program; establish criteria and define
.i ,
the scope of the program; and define the 3, // interface with other work control processes.
,.,', (PRIORITY 2) l *
** 'i NA 4C.13.2.5 Review existing PM tasks and frequency for i s, "
critical equipment. Revise and augment as
., ' required by programmatic selection criteria.
- (PRIORITY 2) f 1 ,
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1 4C.14 ONCE THROUGH STEAM GENERATORS (OTSG's) Latest Amendment Amendnent 1 Complete 4C.14.1.1 Complete Helium Leak Test of both OTSG's. (PRIORITY 1) Complete 4C.14.1.2 Perform Eddy-Current inspection of OTSG's to recommend tubes for plugging. (PRIORITY 1) Cceplete 4C.14.1.3 Plug those tubes identified in 1 and 2 above. (PRIORITY 1) Complete 4C .14.1. 4 Develop program and licensing documents necessary to sleeve tubes in lane region of OTSG's. (PRIORITY 1) Complete 4C.14.1.5 Install sleeves in selected tubes. (PRIORITY 1) l 1 O G S O D-48
APPENDIX E 'l SAMPLE PORTION OF ACTION PLAN e ACTIVITY TRACKING REPORT l 3 July 1986 l i i
O O O
l RECDPeENDATION LIST l N 1 d Coluem Title Descriotion Enlunm Title Descriotion Log No. RRRB identifier defined as: Val /Inval defined as: 15.XXXX section 1$, 12/26/85 Transient IR - Invalid-Redundant Action List Recomunendation 16.XXXX NRC Region V Recomunended Action List ID - Invalid for other reasons 17.XXXX NUREG 0667 and B&W V - Valid 18.XXXX NUREG 1195 19.XXXX Selected Projects 20.XXXX Precurser review 21.XXXX Plant Staff Interviews Disposition Organization assigned by RR8 to 22.XXXX Deteristnistic Failure Consequences Organization investigate and dispose of 23.XXXX BWOG Stop Trip Program proposed recomunendation: 24.XXXX RRR8 Observations CH - Chemistry (these items are not reconsnandations but EM - Electrical Maintenance-but may lead to reconenendations.) propored recomunendation: 2$.XXXX Dept. Managers Reconsnandations IC - I and C Maintenance LI - Licensing NE - Nuclear Engineering Recomunendation Brief description of proposed reconenendations NO - Nuclear Operations RC - Regulatory Compliance RP - Radiation Protection Initiator Person initiating reconsnandation TR Trair.ing TS - Technical Support Disp Eng Dispositioning Engineer
$Y$ System 3 letter identifier of applicitle Due Date Date disposition is due to RRR8 plant system. See AP.3 Enclosure 6.4 from Dtsp. Eng.
(additional designation for this report " tr.cidates disposition (if include a MPS Multiple Plant Systems: MGT - -required) completed. Management NSA - No System Appliable)
$ch Cat Defined as RRR5 recosamendation for:
$U - Startup Category - Area affected by reconsmendation MS - Non Startup Cat May include several areas and are defined: Undeterminded Status Results of RRs review defined as:
NA - NOT APPLICABLE PAG - Recosamendation has been 1 DC - PLANT MDDIFICATIONS forwarded to Performance j MP - MAINTENANCE Analysis Group i LL - LES$0NS LEARNED REC - Recomunendation has been l TR - TRAINING returned for Clarification OP - OPERATIONS AND PROCEDURES DSP - Reconsnendation forwarded to Disposttioning body I EP - EMERGENCY PREPAREDNESS ' 00 - OUALITY AND OUALITY j A$$URANCE l NE - MANAGEMENT EFFECTIVENESS Acted on Date Date recomunendation reviewed by 1 CM - COMMITMENT MANAGEMENT Comunents RRRB CD - CONFIGURATION MANAGEMENT RRR8 Comunents HP - HEALTH PHYSICS RD - RECORD $ AND DATA BASE tot - MATERIAL MANAGEMENT MANAGEMENT
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i Page No. 1 ; 87/81/86 IIRIB IEEDeBOATIM LIST 15I05 !EII) ItX1But EIEUITIIEi STilTISI LMIT GE
********* DE!BER 5,1985 T15tEIBli *+e+++++++ l tai VIL/ DISPGi!TIGl DATE ,
- 10. IEGmeATIIBEi SYS CAT IIML OlGulIZATImi PRIORITY STATLE ACTED i j
l 15.8049 IELML BIEEDIS S1/S2. IITEM 2B, ICS DC V DEIRIIEi 1 PA6 86/02/06 M9uttFACT0151. I 15.0106 PIENIDE APPlWPRIATE SIZE PIETECTIVE ICS DC V DERIIEi 1 PAS 86/02/86 FUIE FOR (3!E R.TDtflTIIEi CWilENT LBftEED L0le CIET Intilldi BEinEEN INTEBMTED CGITIEL SYSTDI l CABIlET H41CS2 NO CGELE H1R1. DEiulE PROPER CIDilD!lflTIGl ETnEEN l M FUSE Ale UPSTIEWI IIEMER. R.50 DEUIE PROPER C00llDD51T131 BEllEDI EIISTIIEi FIEES 48 IPSTIEM PIETECTISL (ACTI M ITDI 3.F.8 CLOSIAE lEPORT). 1 15.8245 IGIFY M PUUR TO EPUE M ICS DC V DEIRDEi 1 PIE 06/03/86 UISTDEi 30A BUS St.T ICS FEEDER BIEMDI WITH A 40A BIEMER (DEDd!ERIIEi D51NIE ICTICE R-0469). (AL7IGl ITEM 3.B.3, CLDIME IEPORT). 15.0247-1 PEIF0lli 101( IEINIIED BY EDI ICS DC V DEIRDEi 1 PAS 06/13/06 R-4159. DEiDEDtIldi D53EE lETICE R-8359 05 DEES M ELECTRICR. SBEllEi StuGS FOR M PGER SUPPLY IDlITORS IN M ICS 130 lell CABIIETS. (ACTISI ITEM 3.F.6, CLOSUIE IIEPORT). 18.8045 EVALUATE M UIE (F A SDELE P0lER ICS DC V BEIRIDEi 1 Pali 86/16/86 EIPPLY ICIITOR FOR ICS, IEi M SIDELE IO(ITOR LDWES M SYSTDI O E-2
pop Ile. 2 57/01/5 m EIDOW W EISILIST WIDE EIII IGE.EM EIEMTildi STIITIGI LIIIT SE occes**ee IEGIE!R 5,1985 TIUBSIEIR *****o**** LE Vit/ DIWEBITI31 DilTE IEL IEIDIEISATIGE SYS DIT IIMt. (Neul!ZATIM pitINITY STimE ACTED (Il HUENE.E TO SIIELE FAILLE. 15.Isf7 OPEMTIWE BGLD IEBIFY ICS 19 V OpBulTIGEi 1 PRS E/16/86 l ApplENIIIATE panmum gg31301 ftlTIR ITBt 1.J 28 3.L1, (1.!RE EpWT. 00186-127). M E-EIElli!ZATII3l STimE F SPECIFIC ICS EINIplGR IS IEEED 10 IEBIFY M ICS IECDBY penmw. 15.8130 A DB58.TY PEEEDIE BELD E ICS Op V OpDulTISEi 1 PRS E5/27/E lEVELDPED 10 plDfl1E RIIDlICE TO M (DEMTORS 5 LIMi F ICS IDER. THIS punmW BRLD E WIT 191 TO ElRE LM N LI5, IE appi.IDeLE, WWE IEB (DOLETED TO M POINT Tmf aCS penes se TB00Um3ES fuE STilBILIZED NS O 15EER DBOETIBITD OPENmNt CDmEL EFOIE ICS TERREHETIE IS CDeWCED. M PMIBLE Flut ICS IGER IESTINIfitBI SRLD STEM (10SIE S1/S2 TtIEMR ICT SEPRlulTELY. 05 /985 IEETDEi 2/18/86). IEBIFY LIB #15.8137 TO E PART F THIS ECDOElelRISL (ACTIGl ITBI 18, CLIBM EPORT, l ECTISI E11. l 15.0063 LIBS F INTEBMTED CIBmEL SYSTBI ICS Ilc, y OPOWITIIBE 1 PRii 85/3B/86 FIEBulES 3GLD DIIECT TIE OF DIERATORS TO PUICE M imEERERIC DLSD VILVE 188 TtNBilEE BYpilBli Vit.VE O E-3
l page No. 3 57/01/M RER IEDeGSRTIGI LIST 1905 !EIl itEIBR ISEMTIldi STRTIGI LIIIT GE
********* IEDEt 5,1985 TlWEIENT m+e++++e i
UB Vit./ DISPOSITI(Bl DATE ;
- m. IEGOGERTIGEi
{ { IBMt. OIEfttIZRTItBI PRIORITY STATLE ACTED MG1R BlITDES IN M 'CUEE' IEEE, M M ltRILIAllY FEED WRTER
- L lue li FLOW CGITML CGmELLERS IN
'NRitNE." IEF01E ATTBSTIldi TD IE-ESTIELISI POS TO M INTEillRTED CGmEL SYSTEIL DEE INTEBRRTED CGmEL SYSTDI P06 IS STABLE, M Vit.YES CAN IE IETUlIED TO M INTEIRRTD CGmEL SYSTEN BY SOINS TO 'le glRL' N W "AUT0*
IESPECTIVE.Y. IEPRIR M INPitpDLY TattliflTED ICS 19 V OPGIRTIGE 1 PIE 86/02/86 15.9287 l WIIE IN M INTEBlWITED CGITIEL SYSTDI CIBIIET. (CGSLETO (ACTIGl ITEN 2R IEOT CIME IEP0lli). ICS 19 V OPEMTIOlE 1 PRii 06/02/86 l 15.0298 AC3ILVE M CRilE F M DIFFDGEE IN M 'AB FRW W.LES F M j TIIE DEUlYS Gl SI 130 S2 (ICS POS SppLY 1011T01t SIITDES) WITH M DESIGN VALLES. (ACTI(Bl ' ITBI 2R MOT CitSE IEP0llT). ICS TR V TMINIIEi 1 PAS 85/30/06 5.0001 IICLLEE " LOSS F INTElilWITED (DGEL SYSTBt" IN M TMINIIEi PIEMiRDL THIS TMINIldi WILL RID M OPEMTORS 11011116i A 'PARTIll. UES F INTEliRRTE CGmEL SYSTBt" AS lELL AS TUTR. LDSS F INTREBMTE CulmEL SYSTEN CASUALTIES. (ITDI 1C, ACTIGl LIST). l i i 4 E-4 O l
page No. 4 87/01/ 3 M ECDIElgrTImp L;gy Neos am unaut asenTDS Smits mIT oE eeeeeeeee IEIBER 5,1985 TWISIBli eseeeeee** LES R / DISplBITItBI IlllTE le. IEIDeGERTIGE SYS CRT IIML (NBulIZATIM PRIORITY STimE ETD R 15.0098 ERINTE M MTLIARY STBul ICS, E V BEltIIE 1 FIE 06/02/86 l leuCIS RVE RSCTIM FM UBS F ABC IllTENTD CDITML SYSTBI E PDER ge enamnT DEWILITY 10 (DfTML MILIARY STBut plEEME NO IlWUIElfT CORIEETIVE ACTIGIIE lE2951RY. (ACTIM ITBI 1.J 28 3.L1 Q.!ElulE lEPORT). 15.0099.A plEVIE A Deli CDML STRTIM IN ICS, E V BeltllEi 1 PAS 86/02/86 M CSITEL EDI Gl H1R110: fLLill IEE InmpflTIC CUMME F M imEERERIC Blff Vlt.VES/RIEDE BYplESVILVES. 15.1999.B plWVIE A IGl CIBmEL STRTISI Ill ICS, E V BERIIEi 1 PIE 85/3B/86 M CDML EDI (Il HAR1 TO: ISS plWVIE 18tt#L QERIE F M fWYS/TBYS. 15.0099.C PNVIE A IGl C9ffEL STRTIM DI ICS, K V' BEltDEi 1 FIS 06/IE/86 M CSML MDI M H1R1 TO: 10 IEE INElBIDE M mmpulTIC QEME F M fmEERERIC Bl30 Vit.YES/RNIBDE BYplai YlLVES IIS II.LSI M OPENmul 10 (Deli M immREltIC DISS VILVES OR TINSDE BYpAlli VfLVES (50 PUEBIT INEIO 14t!LE DITElWIRTED COIML SYSTBt PSElt IS UET. (ACTISi ITBI 3.F.4 fue 5 CulEulE lEPORT, IEFDEIEE BGIldiERIIEi OWEE leTIE R-8357 A/5). 3.8009 Blei STOP-TRIP plEENWI TRIP ICS, 19 V IMDif 1 // E-5
page No. 5 N/01/0 NUB REUBIIBORTID LIST Nuce SEul Mnsut EDGUITIE STRTIM UNIT WE
********* DEEsst 25,1985 TlWBEIDIT **********
LIE WEJ DISPWi!TIGl DRTE
- 16. IEIDOGWRTISEi SYS CRT INWL OMANIZATIGI PRIORITY STRTLE ACTE IEETICII IEUBOGORTION 1911 TlHe9-ICS:
Il910 GENTS IN ICS 115E (2BITIEL CIIE21ITS. 23.00!! BGi STDP-TRIP PIEElWWI TRIP ICS, 19 V IElliT 1 // IEETI(BI IECDOGERTIml 1911 T1H11-ICS: DETagillE IF TE BRID FIEREICY ERIUt CIIIIIIT HAS BEEN DEftBED. 15.0145 L82.14 SOLD BE NEIFID 10 ICS, OP V OPERATI(dei 1 PAS 85/02/86 IEFDEEE ICS IDER RS IELL RS IMI tell IDER IF A CASWLTY PIEXI!ME IS TD E DEVELOPED FOR LOSS (F ICS IDER. A VALIDATION NS VEltIFICATI(Bl IEVIEW SOLD E PDF01 BED Gi RTOS PIEXIDulWL RJIDELIIES EDIESSIls LOSS (F ICS IDER. 23.0008 N ES STOP-TRIP PROBNul TRIP ICS, DP V OPEMTIGEi 1 // IEETIGl IECDeGGRTIWI T1HI6-ICS 1911
- IllplDGENTS TD IERCfDit IUSACK CAP 18ILITY.
23.0012 BiGi STOP-TRIP PIEBWWI TRIP ICS, OP V OPERATIGE 1 // IEETI(M IECDeGERTIWI 1911, TIH12-ICS: DETEllllE IF OPEMTDit 18Ei NAS ISISSARY IIFOIIIITIGI FIDI P80CEDUIES, IISICATORS, ETC., TO DElET LDSS (F Mil POER. ( E-6 9 l
Ilege No. 1 enet/3 RW EIDOEISRTIM LIST . NIDE EIS 18ELEllR IEIElulTDE STilTII218 TIT WE
****ees** IECDER 26,1985 TIUWEIEIE ********+e R / EISN SITIGl IWITE OLDB IEL IECWeGORTIGE . SYS DIT IIMt. OEMIZATIGl Pil!RITY STARE ETED GI ICS, V OPOUITIGEi 2 PA6 05/3B/86 12.IE36.A 030ET A TIERBI EVIE F OP AP.100>-199, PIEEE!E STIUSARS, f88 ffW IIEIRIFY ftS CWUECT ERERS, 70 VERIFY BIDI ETPOINT. WE VERIFIED, M IETPODITS 900 E DEDED flBAllEiT ACTUR. ETPOIIIT lECORDS FOR M DEVIE. M IEVIsl SOLD PROGED ACCORDEi TD A PRIORITIZED LIST F SECTIGEi.
TlelE SECTIGEi IBMLVIIEi SFETY IELATED Ale VITit. EWIPIEIR SOLD l BE DOIE FillST. 23.sgM FDR M TIGIIE lEIEER PIE!BIE ICS, DC V 2 // Il8UT TO M ICS, A IGIFIDITIM 1981 SOLD E IlWLBelTED TO RITWflTIGE.LY IETECT M IIMLID i IlsUT fue SWITDI TO ITS IEDisefult l CIMITBlpflRT. TE CWGPRR. C ESI9Ei FM THIS IGIFICATIGI file DEERIED IN !ECTIGI 3.4 F M IRNICE 1151501T. OE MSIM IS APPLICMLE TO PUIITS 141VDEi ISE THet WE LEADER PIE!BIE SIB 8L PER LOOP fue M ODER DESIGI IS APPLICMLE TO PUuffS islVIIEi SE.Y GE SIS 51L PER LIEP. M E FLOW DSUTS TO M ICS IR E V DEIRIIEi 2 // 23.0001 9eu IE MLETS 25 IGUCD WITH mt E9fIYit.EIE SI9gt.S BLEED Of E Pt#9 STRRE IN N WITH EIT)ER F M TW C0lGPTUlt. MSIDEi PIEENTD IN M EME INX19EllT. COICUMBE WITH M IGUEleff F E FL!nf SIB 5LS WITH EWIRENT SIGNILS, M EXISTHEi (U LIMIT l E-7
- - - - _ _ ~
Petello. 2 j l 07/e1/3 M IEDIGSMIM LIST IIIDE E!Il lu1EM EIMITIIS STRTIM IIIIT GE i
********* IEBet 5,1985 TIBISIEllf i+e*e+ ecee UE VIL/ DISplBITIM MTE
- 10. ECDeGERTItBE STS Dif IIML (NWtlIZATIM PitIORITY STATLE RCTD (Il EMD W E FL21 SELD E IELETD.
ICS, G V BERDE 2 plui 05/27/86 23.0002 NEHi STOP-TRIP PMWutt TitIP MBETIM EGUEIERTIl31 THee-ICS 10l! FM T4ET NS TH2LD, A IEBIFIDITIM BGLD E Il5LBEITG TO RITUEITICILLY DETE:T III DMLID D5UT me SIITDI TD ITS IEneluff CDMBIPAC. M (DCEPTUR. IESIIBl FDit THIS IEBIFICRTIM IS DE50 TIED IN !ECT!!BI 3.2 F M EMEI DlIDelf. 23.0083 BIGi STN-TRIP PNMtl TRIP ICS, K V EleRDEi 2 // EurTI!BI EEDOGERTIM 1911 T H 003-ICS l
- IBDE STNmp Ri FLIN CDUECTIM TD ISDI Fil FLSI RSETIM FIDI M ICS.
23.0005 BEE SHP-TRIP PNMIl TRIP PNBugl ICS, K V BERIlgi 2 // EIRETIIII ECDOG ERTIW1 1911 THE5-!G. MIDE IEUTM FUR 8100LS ACTIWSERDE CIIEUITitY FIDI RPS fue IELII3lTE Ill M ICS. 23.0006 BER STOP-TRIP PENIl TRIP ICS, E V EleRDEi 2 // ElllETIM EIDeBERTISI IIII T H IG-ICS. IELETE Fil 195EMTK IIMCTISI 1D RI IEine FE31 ICS. 23.0007 EXE STOP-TRIP PIEEWWI TRIP ICS, E V BERIlgi 2 // seuCTIM istDeseRTIM ml T H 07-ICS. IEIGE MU LIMITS FIDI ICS. 23.0010 Deli STDP-TRIP PIEBUWI TRIP ICS, K V EIERIldi 2 // E3tCTIGI IEDGEleRTIM let! THING: ICS (DiflEL CIl0 LIT IERIFIDITI(31. IEEEPTIUL ESIM F IEW Tae-CGITIELLEMY-FEEEllRTBt E-8 O\
)
page No. 3 87/91/ 3 M EDIEORT!W LIST IIIDE !ECO III2. ERR IEIEMTIIEi STRTIIBl LMIT GE
*****e+ee DEIBGER 26,1985 TIBIE!Bli ***+oe+e**
LN R1 DII/Ui!TImi DRTE DIT IWAL OhllelIZATICBI PRIORITY STRTUS ACTED 31 E. IEDIEISRTISE SYS CI M IT: M WVIRE FERTINES TO E IIEWWORATED IllTO fIl IBMNED Tave ERER (DmEL IDEIE RE M Ftt.LSillG:
- 1. UIE F R IERIVIITM-PIElpOPTIGEL-INTEiluL C3mu1ER SO THRT FW FLD51 RTE Citt E IEDUCED RS Tave EllROR STRRTS TO DECIERE EYGS ITS HIBEST VII.1E OR FW FLDNIRTE Clut E 11CIERIED RS Tave ERIEIR STARTS IlOEfE!IE FIDI ITS NINilDI R1E. THIS WILL SCRTEN TIE 30 !EIDS UlS TDE fue WILL IEDLII TIE fBOLITLEE 138 PERI!E F TIE (IEILIATI(Il IN Tave RIO IUICTOR POLER.
- 2. INIUT M Tave ERlWR SIBIL DIIECTLY INTO TIE UBP R fue B FlallWITE VEIRE DElWIS CSSRlulTORS
(!McGil SO TifliilE EINER SIBIL WILL CRilE GG.Y TIE 15 TIN CSmEL VALVES TO fEJLET 70 IEDLEE TIE Elul0R. IEFER TO FIIME 7.31 FOR R CDEEPTlfL RAIUUEEleff FGI TIE lEN Tave-CGGELED-BY-FEDullTER CIGIT. M PID CGmE19t (RI.D IEPUEZ M IneER, INTEBlWIL, f38 POSITIVE EMOR NORLES, fue M SISML WTER THE EXISTDE REUIY m WIM.D BE lE-lE!TED TO TIE Funi EMOR COlpARATOR IEELLES IN UDP R fue B FW CIBml[L T15l12. E-9 f
pagelio. 1 57/01/86 M E 300 0RT!WILIST WIDE ! ECD IEELBut IEIENITIIEi STRTIM WIT GE
*******++ IEIBM3t 5,1985 TIUus!Elli ***e+e+ese LIE VIEJ DISPGi1 TIM DATE IEL IECGOEeRTISE SYS CIIT IIME. OEgutIZATIml PRIMITY STRRE ftTED W 15.8296 ESTIE.ISI A PIEVENTIVE 16tillTE188EZ ICS, 19 V OPDIRTISEi 3 FRii 86/M/86 PRIBIE TO DEIX M 01E51T191 F 1811 l M PSER SlPPLY IOIITORIIEi EYSTBI l
Ill M ICS AlO 10ll, IICLIEllEi M TIIE lELAY F S1 RIG S2. (ACTIM ITDI 2R lEIDT CENE lEPORT). I O E-10 O
4 APPENDIX H SAMPLE TEST SPECIFICATIONS The attached Sample Test Specifications are representative of the test specifications which will be used in the test program. j Independent Control of A'h Control Valves -Page H-1 Independent Control or ADV s and TBV's Page H-3 Fire Protection Valve Drain Line Page H-5 Functional Test i 1 i- _ . , _ _ _ ._ _ _ . _ . . _ _ . _ . .. . _ . . _ _ _ _ _ _ . _ . . . _ _ _ _ - _ . _ . . . . _ . _ . , _ _ _ , . _ . _ _ _ _ _ . . _ _ . . . _ , _ , , _ _ _ . _ _ _-
TEST SPECIFICATION Independent Control of AFW Control Valves Revision 0 MODIFICATION NO. NUC ENG COGNIZANT ENGINEER _ NUC OPS DESIGNATED ENGINEER PURPOSE: To verify proper operation of the following:
- 1. Dual steam generator level indicator on H2PS.
- 2. Preset automatic positioning of the Auxiliary Feedwater (AFW) control valves on loss of ICS DC power.
- 3. Two manual control stations for the AFW control valves on H2PS.
REFERENCES:
ECN R-0357A and associated drawings. General Calibration Procedures I-011. O
- Calculation for preset AFW valve bias for Loss o' lis Power -
Z-FHS-IO105. PREREQUISITES:
- 1. Construction complete and turned over to Startup with any exceptions duly noted and recorded.
- 2. Ensure simulated loss of ICS does not adversely affect the plant.
- 3. Auxiliary Feedwater lined up to permit stroking of the AFW control valves.
TEST METHOD: I&C Maintenance Department to verify the following:
- 1. I.astallation and operation of Solenoid Valves FY-20527C and FY-20528C.
- 2. Installation and operation of Auxiliary Feedwater Valve Manual Control Panels HC-20527 and HC-20528 (located in H2PS).
- 3. Installation and operation of I/PS FY-205288.
l
- 4. Installation and operation of Steam Generator A and B Startup Level Indicators LI-20503B and LI-205048 respectively.
H-1
I The above will be verified by the following methods:
- 1. I&C Maintenance Department to verify installation and operation of instrument loops by calibe ating instrument loops using loop form per '
General Calibration Procedure I-Oll. ; 1
- 2. With ICS power avallanle, and Manual Control Stations in automatic, verify that Scienoli Valves FY-20527C and FY-205?8C are de-energized and verify the following: ;
- a. Red indicatot lights above L&N Manual Controlleia are not lit.
- b. AFH Control Valves receive pneumatic control signals from ICS via E/P's by using existing Bailey Hand / Auto stations on Panel HlCO.
- 3. Simulate a Loss of ICS Power then verify that Solenoid Valves FY-20527C and FY-20528C are energized and red indicator lights above L&N Controllers are lit. Then perform following:
- a. With L&N Controllers (HC-20527 and HC-20528) in automatic, adjust the preset bias potentiometers to provide the required initial Aux 111ary Feedwater Control Valve position for loss of ICS DC power. The valve position will be determined from the calculated AFH flow needed. Record the valve position at both AFH Valves L&N Manual Controllers (HC-20527 and HC-20528) indicator,
- b. Place L&N Controllers in manual and calibrate surb that 0 to 100 percent indication on the Manual Control meter w esponds to O to 100 percent Auxiliary Feedwater Control Valve st, travel. Verify "bumpless" transfer from Auto to Manual for tN s'.N Manual Controls.
- 4. Place the Manual Controlle s in "AUT0" and remove simulated Loss of ICS Power and verify FY 2052.7C and FY-20528C de- energize and the red indicator lights go off.
- a. Place L&N Controllers in " MANUAL" and verify manual control as in 3.b. Insure red indicator lights are lit.
- b. Place L&N Controllers in "AUT0" and verify return to normal ICS control by modulating AFH valves using the Bailey Hand / Auto Stations on HlRC
- 5. With Manual Controllers in "AUT0" and no simulated Loss of ICS Power verify normal ICS Control as in 4.b for each of the following conditions.
- a. Removal of AC power to the 24 VDC power supply for HC-20527 and HC-20528.
- b. Removal of all AC power from SIN 1-1 to H2PS feeding the new equipment installed by this modification.
ACCEPTANCE CRITERIA: Equipment tested and results meet criteria under Test Methods. H-2
TEST SPECIFICATION. Independent Control of ADV's and TBV's O () Revision 0 MODIFICATION NO , _ , _ _ NUC ENG COGNI7ANi ENGINEER NUC OPS DESIGN".'M ENGINEER START-UP C00RDINMOR PURPOSE: To verify proper operation of the following:
- 1. Selector switches for the Atmospheric Dump Valves (ADV's) and Turbine Bypass Valves (TBV's) on the HlRI panel.
- 2. Auxillary relay contacts in the ICS.
REFERENCES:
- 1. ECN R-03578 and associated drawings.
PREREQUISITES: j O 1.- Construction complete and turned over to Start-up with any exceptions duly noted and recorded.
- 2. Ensure simulated loss of ICS does not adversely affect the plant.
i 3. Plant in cold shutdown and secondary side of Steam Generators depressurized or in a configuration to permit testing TEST METHOD: I&C Maintenance Department to verify the following: l ( l. Installation and operation of the ADV and TBV Override Switches, HS-20562C and HS-205618 respectively located in HlRI. l 2. Wiring to ICS relay 86/ICS-PSM contacts. l The above will be verified'by the following methods: ADVs 1 1. With both hand switches HS-20562A (located on Shutdown Panel H250) and HS-205628 (located on Atmospheric Dump Valves Manual Control Panel) in
" Auto" and HS-20562C (located on HlRI) in " Normal" verify:
O H-3 i ~,
- a. Solenoid Valves PY-20562D, PY-20562E, PY-20562F, PY-20571D, PY-20S71E and PY-20571F are energized.
- 1. Simulate a Loss of ICS Power and verify that the solenoid valves in 1.a. above de-energize. Verify ADV's are closed.
- 2. Place the ADV Override Switch HS-20562C in " Disable" and verify that the solenoid valves in 1.a. ebove re-energize.
- 3. Place the ADV Over:1de Switch HS-20562C in " Normal" and remove the simulated Los's r f ICS Power Condition, verify that the solenoid volves in 1.a are energized.
- 4. Place the A: . Ovei ride Switch HS-20562C in "Close", verify that solen.') valves in 1.a. are deenergized. Verify ADV's are closed. l TBVs
- 2. With TBV hand switch HS-20561A (located on H2SD) " Auto" and HS-20561B (located on HlRI) in " Normal", verify:
- a. Solenoid valves PY-20561A, PY-20563A, PY-20564A, PY-20566A are energized:
- 1. Simulate a Loss of ICS Power and verify that the solenoid valves of 2.a. are deenergized. Verify TBV's are closed.
- 2. Place the TBV Override switch HS-20561B in " Disable" and verify that the solenoid valves of 2.a. above re-energize.
- 3. Place the TBV Override Switch HS-205618 in " Normal" and remove the simulated Loss of ICS Power Condition, verify that the solenoid valves in 2.a are energized.
- 4. Place the TBV Override Switch HS-205618 in "Close", verify that solenoid valves in 2.a. are deenergized. Verify TBV's are closed.
ACCEPTANCE CRITERIA:
- 1. Equipment tested and results meet criteria under Test Methods.
O H-4
APPEMOIX I
;cI No .11, lev. 1 July 11, IN6 Attach:ent No. 3
'h\ Page 1 of 6
.. (
\ PLA.'C STAFF INTIRVII'4 1.0 Purpose The interview process shall identify any previously unidentified, but "known" proble=s which =ay impact the safe or reliable operation of systems or components of the plant. This procedure is to provide guidelines for the interview process.
2.0 Score Interviews will be conducted with, a:ong others, plant operators, operations tech support engineers, plant engineers and technicians, training instructors, design engineers and quality depart:ent personnel. These personnel will be encouraged to identify systes, component , or operational related proble=s, or concer:s which they are aware of and any recommendations to-resolve them. It is intended to interview personnel fres each plant organizational group. Volunteers will be requested. All volunteers will be interviewed. If insufficient personnel volunteer, the interview progras coordinator will perfor= a randos selection of i terviewees. e ,j/ W The si 1: = nu:bar of person:nel to be interviewed are derived froc (A*P 3, which is based on MIL-STD-105D. Other interviewees =ay be adfed at (k tne discretion of the Interview Program Coordinator (Inclosure 1), 1 3.0 Reseensibilities The responsibility for the conduct and docu=entation of the interviews rests with the Interview Program Coordinator. The interview teama are responsible for conducting and docusenting the interviews. The Interview Program Coordinator is responsible for compiling the results of the interviews. The Interview Program Coordinator shall ensure that interviews are thorough and are co:pleted is a ci:ely canner. The Interview Program Coordinator is responsible to the Perfor:acce Analysis Group for the conduct of this progras. I l Initially four interview tea:s will be utilized consisting of hu=an
- factors engineers and other knowledgeable plant representatives. These j interviewers will be responsible for conducting the interviews and I cc= piling the data. The interviewers are accountable to the Interview
( Progra: Coordinator for tabulation of interview reco==endations. Interviewers can be added or deleted as dee:ed necessary by the Interview Program Coordinator. 1-1 l
.s. . . ..o...,....
July 1*., I? H A:.ach:c:: 3
?sga 2 of 6 N 0
O'
.?:ocedure The nu=ber of 1::erviews that will be required :o e:sure adequa:e coverage are listed 'in helosure 1. hese are de =ini::= su=be:
required, but if the interviewers, with concur:e:ce of de Interview Project Coordinator, decer:hes : ore interviews are desirable, they =ay conduct them. The interviews say be conducted with g: cup of person =el to provide a synergistic approach, or on an individual basis, c a cc:bina: ion of both. The app cach used will be de:e::1:ed by :he interviewers with ; concurrence of the Interview P:cgra: Coordinator. The interview will be acco:plished by requesti:g the area superviso: to provide the case (s) of personnel that have volunteered :o be interviewed. If insufficient pe:sonnel voluntee: :o =eet the require =ents of Ecclosure 1, the Interview Progras Coordisator shall select additions 1 persennel to be is:erviewed.
.4.11 nuclear =anagers and their superistendents shall also be approached for interviews.
All concerns, ;;obless or :ecoc=endations identified duri:g de m interview shall be reco::ed on the Personnel Interviev For=, Inclosure )
- 2. A copy of the questionnaire used for the interviews is included "
/
the progras plan. All concerns, proble=s and thei respec:ive recoc=endations (if any) fro: the Perscnnel Interviev Fo:= will be co= piled into one list by de interview tea =. Redundant concerns, proble=s or reco =e:ations will be consolidated into cue statenest to reduce the nu=ber of concerns. If redundant concerns are consclidated, the total cusber of people with that concern will be noted by placing that nu=ber in paren:hesis at the end of the consolidated concern. The ec= piled list of concerss/reco::endations vill be forwarded to the Reco==e:da: ion, Review and Resolution Soard for Processi:3 The I :arview P:cgra= Coordina:or shall ':e considered the ici:ia:or on de Recoscendation/Resolucios for:s. O e
- .C No.12, Eev. 1 July 11, 1966 2- ac'- e- ~'
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Af ter the concerns /reconnendaticus have been dispositioned, it will be , - the responsibility of the Interview Progran Ccordinator to infors the interviewees of the disposition of their cencerns. For group interviews, ev.eryone in the g: cup should be notified. 5.0 Budeet and Plan- - The Interview Progran Coordinator assesses the resources required to conplace the tasks as prescribed and shall present the Perfernance Analysis Group with a plan that identifies the hunan and other resources required to execute the tasks and the approxinate cine required to conplete all tasks. The PAG shall review the plan for .conpleteness and for app:cach and shall direct approval to proceed or revisicus. The PAG shall ensure that resources are available for the approved plan. 6.0 Interview Guidelines The interviewers shall ensure that interviews specifically address the following plant systens/ areas: Reactor Coolant Systen, Reactor Auxiliary Systens, Secondary Systens, Plant Support Systens, and Instrunentation and Control Systens and electrical distributien, as they
' 75 apply to the interviewee's experience and knowledge.
' 's The purpose of the interview should be kept obvicus during the interview: the identification of potential or known probleus within systens, conpenents, or operational guidance that could inpact reliable plant operation. Renenbar that the main point of the interview is to learn what others believe nay be potential or actual problens. The Laterview tean nust stimulate the interviewee into describing total experiences. The interviewers shall develop a questionnaire based upon the background of the interviewee. This questionnaire shall be used during the interview to help the interviewee recall previous probless or concerns they or others have had. l > l \
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, .. - . .. .w M y 11, 1966 Ac:ach:en: 3 .s su , ,-
Page I. :f i / v.2*Dit?. NUM3E?. OF PI?.SONNI; 20 31 IN!.".VII*aT.D . Plan: C:erators Shif: Supervisors - 3 Senior Control F.com Operators - 3 Reactor Operators - 5 Auxiliary Operators - 3 Iqui;:ent Attendants and Poved Plan: Helpers - 8 Ceerations Tech Sueco:: Tech Support h aineers '5
. Shift Tech:ical Mvisors - 2 Scheduling - 2 Plant Maintenance I&C Technicians - 5 Isc hgineering - 2 Co=puter Tech =icians - 3 Ilectrical .v aistenance - 8 Ilectrical Engineering - 3 Ilec::1 cal .Techs - 3
.v.echanical Maintenance - 6 Mechanical Ingineers - 3 SMLO Securi:r - I.
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SCI No.12, F.ev. 1 July 11, 1356 A::achten: 3 7% h elesure 1 Page 3 of 6 s MINIML?. NL'M3ER OF PERSONNEL TO SE INTERVII"E3 NPS - 20 Licensing - 4 Training Instructors Operations Training - 5 Maintenance Training - 3 General Sployee Training - 2 Nuclear Engineering Electrical - 8 I&C - 8 - Mechanical - 8 Civil - 2 Project haineering - 3 Quality Department QA Inspectors - 3 QC Inspectors - 3 QE Inspecccrs - 2 Corporate QA - l Rad Protection Senior Chen-Rad Assistant - 2 EP Tech - 5 Chenistrv Senior Chen-Rad Assistant - 1 Chen '.-2 Tech - 5 'k t-s
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Page 6 of 5 I:cIssu:e 2 PERSCNNZI. D;TZ?.VII'4 FCRM 3o . ITEM DESCII!TICN N
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Syscas Desiges:ics PLANT SAFI!Y/RILIA3IL,I"Y - P105 LIM /CCNCZ2N/CONSEQUE:CIS e3 RICOS.ZNDED SOLUTION (OPTIONAL) l l
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