ML20206T888
| ML20206T888 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 07/31/1986 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20206T873 | List: |
| References | |
| NUDOCS 8607080248 | |
| Download: ML20206T888 (346) | |
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.p TABLE OF CONTENTS SECTION NO. TITLE PAGE NO.
4 1.0 Introduction & Program Overview 1-1 2.0 Management of the Action Plan 2-1 l
3.0 Performance Improvements Underway Prior to the 12/26 Event 3-1 :
- 4.0 Restart and Performance Improvement Action i i
Plan 4-1 4A Systematic Assessment Program 4A-1 )
48 Management, Operations, and Administrative !
-Process Improvement 48-1 4C Plant Modifications and Maintenance Improvements 4C-1 i
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40 System Review and Testing Program 40-1 a Appendix A District Board of Director's Policy Statement on Performance Improvement 4
at Rancho Seco A-1 Appendix B Specific District Responses to NUREG-1195 Findings B-1 Appendix C Cross Reference Action Plan to NRC Open Items C-1 Appendix 0 The District's Assessment and Comparison of Test Programs 0-1 Appendix E Sample Portion of Action Plan Activity Tracking Report E-1 Appendix F Schedule Appendix G Rancho Seco Restart Implementation Organization System Status Report G-1 Appendix H Sample Test Specifications H-1 d
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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 similar 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 (SMUD) 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 i
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 iQ projected electric demands in l
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the SMUD service area, and the benefits to be derived from achieving a higher level of performance, justify the additional investment l
required to achieve the designed results. They also realized that reliability improvement is closely coupled with safety improvement l which has always been a first priority.
Before actions associated with this decision to improve performance could take effect, the importance of the program was reinforced by a number of undesirable operating experiences in 1985. The most significant of these events occurred on December 26, 1985, when a loss of power to the plants integrated control system led to a plant overcooling. The cooldown rate specified to limit the stresses induced in reactor systems heavy metal components was exceeded.
While subsequent analyses determined that no serious stresses were induced, the significant potential of this event is not to be 1
understated.
O 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 i has documented their findings in NUREG-1195. The conclusions and recommendations identified and contained in NUREG-1195 are consistent with those reached by the District. In general, these findings are:
- 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.
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- 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.
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- c. Maintenance program deficiencies contributed to the inability to i
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mitigate the severity of the cooldown transient.
- d. Non-compliance with existing procedures contributed to the overcooling event and caused additional unnecessary complications.
- e. Manufacturing defects in the electrical terminations of particular control cabinets (ICS) initiated the event.
.1 Post Event Review The post event reviews by the District 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 1-3 t
- IC3 Controlled Valves j - Makeup pump i - 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 i
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- Health Physics Emergency Preparedness Based on these findings, SMUD has modified, expanded, and accelerated the implementation of its performance improvement g action plan.
The objectives of the Action Plan are to:
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- 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 compliance with license requirements
- 5. Minimize the need for operator actions outside the control room. j
() 6. Improve the reliability and availability of the plant 1-4 I
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Conditions which are outside the window may involve challenge to . safety systems and result in events such as over cooling, under cooling and loss of subcooling.
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i 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 j performance of the Rancho Seco power station are identified 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, I
effective, and timely manner consistent with their i importance to safety and reliability of plant operations.
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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.
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O O O ISSUE EVALUATION &
DISPOSITION PROCESS INPUT MANAGEMENT MPLEMENTATION CLOSURE Dept. figrs. Restart Testing Hardware & l1odifications &
Programmatic , f1aintenance Reconunendations Improvements 1r t1anagement Operations & @
m if if L Administrative Verification
[1anagement Performance Process Validation Process Analysis Improvements Filing Review Group
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Tracking Systematic Recommendations SYSTEl1S REVIEW Assessment Review and ystems Engr.
p Functional Reqmts.
Resolution Board Test Reqmts.
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'N - 1.1 INPUT PROCESS (ISSUE IDENTIFICATION)'
The process, to assure that the review of the various areas which can impact plant performance (management and management processes, plant
, design, operations and operating procedures, maintenance, training, and other support activities) is adequate to identify any deficiencies, consists of activities which review these impact areas from four perspectives. Figure 1-3 graphically illustrates this perspective relationship of: (1) the top down department managers hardware and programmatic recommendations; (2) the bottom-up t
systematic assessment program elements; (3) the management process review; (4) and the system review and test program. Each of these perspectives has advantages and disadvantages relative to its O 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 i
December 26, 1985 event. A description of these Issue Identification Action Plan elements is as follows:
.1 Department Managers Hardware and Programmatic Recommendations An assessment of the plant design, management, operations, and administrative system deficiencies was conducted based on the functional organizations' knowledge and existing documents of O previous evaluations by others. The following important sources were used as input to this assessment.
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O Y hO Precursor Review - Let's, Maintenance Program
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- 1> oh llaining Program Deterministic yn f h p&
Failure Consequence
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Lessons leamed Program
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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 I (see Section 48) and the plant modification action plans (see Section 4C).
- .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.
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'N The near-term and long-term action plans to improve management effectiveness which resulted from these actions are contained in the management, operations,-'and administrative process a
improvement-action plans (see Section;48). The MPRG team will-facilitate the implementation of all management effectiveness action plans.
.3 Systematic Assessment Process (OCI-12) 4 A comprehensive broad-scope, systematic assessment program Plant Performance and Management Improvement Program (PP&MIP). This program was developed and is being implemented to perform a
! detailed review of the plant design and experience as well as 1
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 i
program. Specific input areas include:
- a. Precursor Reviews
- b. Plant Staff Interviews-
- c. Deterministic Failure Consequence Reviews r
- d. B&W Owners Group Step Trip Program
- e. December 26 Event and NUREG-1195 items
- f. Selected projects ,
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. 4- System Review and Test Program (SRTP).
The system review and. test program is a key element in the ,
, implementation process. The systems engineers implementing this program are in a position to obtain an overview perspective of the issues and recommendations. This overview perspective 4
combined with the implementation of the system review process-may lead to the identification of additional issues and recommendations.
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O 1.2 ISSUE EVALUATION AND DISPOSITION PROCESS MANAG'EMENT The process to address and manage the resolution of issues is described in detail in a special Plant Procedure, QCI-12. An i overview of the steps of this process as well as a description of the RRRB, PAG, the prioritization process and the oversite role of the independent review group is as follows.
.1 Process Overview Issues, along with recommended solutions, are identified by the special task or input groups.
1 These issues and recommendations are sent to the Review,
, Recommendation, and Resolution Board (RRRB) for validation i
and acceptance.
1 Recommendation involving systems under review are sent to
, the system engineer and others are sent directly to the Performance Analysis Group (PAG) (made up of the Nuclear Department Managers). All recormendations, both valid and invalid, are dispositioned to the. satisfaction of the PAG.
This group confirms the disposition of the priorities and i 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.
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] Once implementation action on valid recommendations is completed a close out record is prepared and sent to QA for verification of content and validation that' the implemented action is consistent with intent of the-issue and
- recommendation.
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.2 The Recommendation, Review, and Resolution Board (RRRB) is a nine member multidiciplined group of individuals with nuclear experience and training drawn from SMUD, another utility with a B&W NSSS, NSSS Vendor, and the l ? ant Architect Engineering firm.
As described in QCI-12, the RRRB is to determine the " validity" 1.e., the correctness and uniqueness, of each received recommendation. They consider only the technical merits, not the cost, time, or resources available. Once they determine a recommendation, it is passed to the System Engineer or the Performance Analysis Group (PAG) for evaluation and disposition.
.3 The PAG membership is composed of the Managers of the Nuclear Departments or their designees. This group has knowledge of the competing priorities, needs, commitments, and resources available to resolve each recommended action. The member nuclear departments are:
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' Nuclear Projects, Chairman Nuclear Licensing .
Nuclear Quality Nuclear Plant Nuclear Engineering Nuclear Training In addition to receiving input from the RRRB, the PAG also i receives inputs from the Management Process Review Group, the Systems Engineers for System Review and test issues, and 4 .
recommendations from the department managers themselv.es. All
- input is evaluated against 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.
h Implementation of the dispositioned recommendations after
! approval by the Assistant General Manager (Nuclear), falls to the line organization 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 operations.
.4 Activity Prioritization Criteria The criteria by which activities to resolve issues are prioritized and placed in the appropriate disposition period is as follows:
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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 window
- b. assure compliance with technical specifications i c. minimize the need for-Operator Action outside the control room within the first ten minutes of an event.
Near Term - Actions to be-initiated as promptly as practicable, schedule developed, resources ass'gned and maintained until completed which will:
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- a. enhance the ability to remain in the post trip window (e.g., auto action vs. operator)
, b. reduce reactor trips
- c. reduce challenges to safety systems i
- 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' performance objectives.
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' l.3 IMPLEMENTATION OF ACTION PLAN ACTIVITIES The majority of the actions to evaluate the impact of the December 1985 event and to preclude this event from recurring 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.
- These implementation action plans, while broad, comprehensive, and likely to address any significant deficiencies, cannot be finalized until the evaluations of the systematic assessment program (PP&MIP) have been completed. The two areas of implementation are; a) plant maintenance and modification actions and b) management, operations,
- and administrative process improvement actions. Each of these areas is described below. !
.1 Modifications and Maintenance Improvement Actions A description of the specific major modifications and
- maintenance improvements is provided in Section 4C along with the prioritization for each. Most of these items include more I
than a single recommendation or disposition to accomplish their implementation. As such, Appendix E is provided, which lists by.
system, all of the valid recommendations developed and processed to date. It is expected that new items will continue to be identified and processed through the existing systematic review program until the end of August 1986.
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.2 Management, Operations, and Administrative Process Improvement l
Actions The major management and administrative process area's identified
} for improvements are described w'ith the associated prioritized t i dispositions in Section 48. Appendix E lists all of the valid 3 ;
4 programmatic improvement actions associated with management, I operations, and the administrative processes. )
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1.4 SYSTEM. REVIEW AND TEST PROGRAM v
There are two system review programs to be conducted as part of this action plan.
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 a review of previous tests conducted. This action will t
provide additional assurance that these systems have retained, or have adequate analysis to justify differences, their FSAR functional basis, and 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.
.1 Systems Review and Test Program The System Review and test program, which is modeled after Davis-Beese, is a key element in the' implementation process; the systems engineers may also identify issues as they implement the system review process.
The systems engineer program being implemented at Rancho Seco is modeled after the INPO GOOD PRACTICE . In this program the system engineer is responsible, among other things for: (1) the 1-19 f
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J Q development of system functional requirements; (2) assuring that coordinated and effective disposition of action plan recommendation system deficiencies occurs; (3) taking the necessary steps to assure the adequacy of previous system testing; and (4) to develop and maintain a set of test requirements which assure the material readiness and operability of each system. To assure that this can be accomplished in an efficient and effective manner all dispositioned recommendations are not only sent to the dispositioning department for action l but are also sent to the system engineer.
.2 System Functional Review
[~'g The systems which have been identified as dominant causes for severe and complex post trip plant transients will undergo a more extensive system review than that identified.
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 1-20
i which are critical to assure that,'immediately after a reactor trip, the transient remains in the post-trip window (see Figure 1-3). 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:
- Main Feedwater-System
- Auxiliary Feedwater System l - ICS/NNI
- Pressure Control functions of the Main. Steam System
- Instrument Air This comprehensive review of these five selected systems will be initiated prior to restart and will be completed i
prior to coming out of the cycle 8 refueling outage.-
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f 1.S ISSUE RESOLUTION CLOSE00T F.
- A formal process will be implemented to assure the effective and-
, complete closecut.of action plan' items. This is accomplished through
{ the Quality Assurance department which is charged with the i
responsibility to verify that actions were taken in accordance with l
the plan and existing procedures and to validate that the actions
} taken meet the intent of the original recommendation and resolve the i,
j original issue.
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1.6 INDEPENDENT PROGRAM OVERSIGHT
, Independent oversite of the Action Plan is provided by an independent review group (IRG). Tflis group consists of senior persons with i
significant experience in the management and oversite of nuclear
- power plant operations, design, and regulations. j 4
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4 1.7 DOCUMENT PURPOSE 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.
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.1 ' The body of the report is organized as follows:
- a. Section 2.0 Hanagement of the Action Plan - describes Management processes and features of the program to assure
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that appropriate actions are being taken to effectively, 4
efficiently, and thoroughly identify, prioritize, control, and implement changes to resolve the type of deficiencies which contribution to the December 26 event.and the poor performance record of Rancho Seco. This includes:
The independent review group-makeup and role l
The Restart and Implementation Organization (RIO) 4 The Action Plan Tracking, Reporting and Close Out i process The Action Plan Adjustment process The Transition Plan
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- 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 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 Action Plans.
- 3) The Plant Modifications Action Plans.
- 4) The system review and test program.
.2 The appendices of the report are organized as follows:
- a. District Board of Directors Performance Improvement Policy Statement.
- b. The~ District's Response to the NUREG 1195 Findings and Conclusions.
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![ c. Cross Reference of NRC Open Items to' Action Plan Sections. i
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- d. Test Program Comparison Matrix. ;
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- e. Sample Action Plan Activity.
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- g. Action Plan Schedules. :
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l h. Example System Review and Test Report.
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3 o l.8 CONCLUSION J
As the District management team shaped this program,'it became clea'r 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 a 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 tne 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 cf activities important to restart and performance improva % t U Rancho Seco have been completed. The SMUD team is $.att.a e/ to bringing about this performance improvement in an orderly, safe, and effective manner and i believes this Action Plan to be the vehicle for such change.
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s 2.0 MANAGEMENT OF THE ACTION PLAN a
This section describes the actions the District has taken and 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 temporary Restart and Implementation Organization (RIO).
c) Es, tab 11shed 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.,s m ..+. -
incorporates appropriate features of the Action Plan into the line organization to achieve and maintain a high level of plant performance. -
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2.1 INDEPENDENT REVIEW GROUP (IRG)
.1 Purpose An IRG, made up of Messrs. John Jackson, Richard De Young, James ,
O'Hanlon, and Arthur Gehr, has been established to periodically l provide the District's General Manager and Assistant 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.
r.,,,..-,,,...,
John Jackson -,cManyryears,4s _a 2 quality aexpertuin designvwsetoi-uv.wmw+*
construction, operation and training, as well as the analysis-and oversight of quality functions. '
i 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.
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i Arthur Gehr - A recognized expert in nuclear law and regulation i y ,j who played a significant role in the development of the nuclear programs at Commonwealth Edison and Ari' zona Public Service Company.
.2 Mission The mission of the IRG is to assure the General Manager 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.
.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 Washington
- 3. Rancho Seco staff personnel
- 4. Others Carl Andognini (Consultant to the Board) l l
4 J. Mattimoe (Former SMUD General Manager) m l
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5 E. Wilkinson (Former President INPO)
INP0 (all key managers)
The SMUD Board of Directors Attend key meetings .
i 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.
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Engineering
- Quality Assurance Purchasing Maintenance i
- Li cens i ng. (USAR) . . .-~... L - -Trai ni ng -~m.., *: . . o . 4 ~,4. +- .
1 Operations
- Regulation-Review Technical Presentations IIT Report SALP Report history LRS INPO PAT Reports Participate in Plant Tours i
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Review Organizational Issues
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Organization Functional responsibilities i i Personnel ,
l Delegation of authority i
Perform reviews of key operating and administrative procedures.
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. 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?
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,y ,. , -, - - - y,- , . ,yp. -.~--.w--#,-,...y-, ,m v. v e.a r , ,-,,,, yy ,, , ---. ,, y y .,,,,g y- ---g-- wr ve
i Are the priority of action criteria appropriate to assure j that.those actions taken before restart are sufficient to reasonably assure safe operati'on? i 3
l, Is the executive direction and the management plan .,
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! appropriate to provide proper implementation of the Action i
Plan?
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. _ . _ . . _ . . _ _ _ _ _ . . _ , _ _ _ , _ . . . . . _ _ . _ _ _ . , . _ . - _ . . . . . . . _ _ . . _ . - _ _ _ _ _ . . - . . . . _ _ _ . _ , . . _ _ . . _ _ _ . _ . ~ . . ~ . _ . . , _ , _ . _ _ . . , _ _ . . _ , .
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 organization will remain in effect until disbanded by the Assistant General
. Manager, Nuclear.
.1 Responsibilities The Assistant 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 content and the schedule.for
,. impiementation....He has the.: author 1.tyr toabolish:.the. temporary.wwun* -eu -
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 AGM, 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 b
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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 Test Program Director and Program Ofrector. He provides functional direction for restart and improvement activities to the Outage Manager and the Nuclear Department Managers.
Administratively, the Nuclear Department Manager's reporting relationship remains unchanged.
The RIM has authorization to issue instructions as needed to l further define the Restart and Implementation Organization or to control interfaces to effect the action plan.
l The Outage Manager -is- responsible-to the-RIM for execution of '-~~-~~ ' ~ ~ ~ ~ ;
the physical work items required for restart. The Outage
! Manager also provides. scheduling.for the entire Restart and -
Implementation Organization. He has the authority to rearrange sequence of work within limits of the plant technical 1
specification requirements, plant operating procedures and"---
i schedule commitments. Changes which would require extension of schedule commitments, or-plant procedure changes, shall only be 1 made with the concurrence of the RIM, and for tha latter case, i .
the Manager, Nuclear Plant. ,
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The Test
- Program Director is responsible to the RIM for l developing and implementing the test program identified in the action plan. This includes: I'dentification 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 (wlth the Outage Organization and Nuclear Operations) the performance of surveillance tests; working with the Outage Manager in the development of a detailed test schedule; and evaluation 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 j Procedures", or AP. 303, " Surveillance Program".
The Program Director is responsible to the RIM for oversight o n .,.3 m ,-# # ~- , and-- coordination of,tthe..implementationrofathose (progr.ammaticeurme ram v commitments-in the--action plan. -In this capacity,-te
- works with"-~ ~ ~ ~~
the Nuclear Department Managers or their designees to ensure that the proper.. interfaces. and priorities are maintained and - -
that resources are used effectively. This position will work with Nuclear Department Managers and the outage organization to develop a detailed -schedtrie.-In -addition ~to thoseTommitments'-' ""~ ' ~
contained in the action plan, the Program Director shall also ensure implementation of commitments contained in the coordinated commitment list required for restart.
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I The Manager, Nuclear Plant is responsible to the RIM for providing the required operations, maintenance, health physics and technical support to accomplish the scheduled activities.
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- This includes timely review of procedures and test results by l i
the PRC and MSRC. He is responsible for the preparation of .
i reports as required by the RIM. In addition, he is responsible for assuring that programmatic commitments in the action plan
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are not only met in a timely fashion, but are coordinated through the Program Director to ensure consistency within the i
overall programmatic improvement plan.
The Manager, Nuclear Engineering is responsible to the RIM for providing the required engineering and construction support for i design modifications, engineering studies and other engineering i
- support required to accomplish the schedule activities. In 1
! add l t i on , . he. I s respons i bel ar. forums suring@a tuprogrammatWmm b9N v"' 'm
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commitments-in-the action plarrare not only7 net-1n a timely ~~~ ~
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fashion, but are coordinated through the Program Director to i
_ . ensure consistency wlthin the overaH-programmatic improvement--- - ~ '
i plan. He is responsible for the preparation of reports as j requested by the RIM.
The Manager, Nuclear Projects, in the role of Performance Improvement Manager, is responsible to the RIM for the activities of the Management Process. Review Group and those I'
review activities conducted under the auspices of QCI-12. He is
- also responsible for preparation of reports as required by the RIM.
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The Manager, Site QA is responsible to the RIM for providing l
,9 y 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. Corporate QA
' will retain the independent audit function. '
4 The Manager, Nuclear Training is responsible to the RIH for providing the required training support to accomplish the scheduled activities. This includes the identification of needs, in concert with appropriate line managers;. development of i
! training materials; presentation, evaluation and documentation i
, of the training. In addition, he is also responsible for assuring that programmatic commitments in the action plan are w,-am.v ~. m~c~ . coordinated through-thecProgram Btrectoritoessserencensistencyst**cmm'5 i
- within the overall programmatic-improvement plan:- -- ' ' ' ' '
i l . . . The Manaaet,. Nuclear Licensing-is.' responsible 4o-the-RIM -for ~ - ~~ - -
providing the required licensing and engineering planning t
l support for the scheduled activities. This includes the
! establishment of licensing strategies,- in concert with the line i managers; preparation of submittals in a timely manner, I
r coordination of meetings with NRC and other agencies; and j support for-Training in preparation and presentation of l emergency preparedness training. In addition,.he is responsible i
for assuring that programmatic commitments in the action plan j are coordinated through the Program Director to ensure l i consistency within the overall programmatic improvement plan.
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.2 Qualifications
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Assistant General Manager Nuclear - John E. Ward Mr. Hard is an experienced chief executive-level, ,
nuclear-expertenced manager with 34 years proven performance in planning, directing, and analyzing complex operational and engineering projects and organizations. He has completed naval reactors tral,ntng and has advanced s
degrees in nuclear physics. He is a recognized expert in the area of the utility industry regulatory processes. Mr.
Ward 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-
. ;ls,a registered 3professionaleengineerrin therfieldscoftaio22:rvic ,:ce : ~t Mechanical and-Nuclear Engineering in-the-State of ------- --~~~ ~ ~ '"~ "
Restart / Implementation Manager - Dan C. Poole l
1 Mr. Poole-has-25 years of experience-in-the nuclear ~ field. ~--- ~-~
i This includes serving as Plant Manager cf the Crystal River Nuclear Plant, with responsibility for all aspects of operation, maintenance, and technical support. In
. addition, he has also served in the capacity of Assistant Manager of Operations and Maintenance of the Callaway s
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Plant, and the capacities of Superintendent Operations, Superintendent Technical Support, and Training Supervisor.
at the J. M. Farley Nuclear Plant.
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b Test Director - Jim Field ,
i 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 i
recently headed up the group which performed the Deterministic Failure Consequence Analysis described in Section 4.A.
4 Program Director - T. C. Lutkehaus 1 ,
~ s' h,m w w , . .- Mr. Lutkehaus_has,overmlkyears of nucleart poweraplantucow.arretwanc
.. - - - - - . - experience, which-includes-.12 years experience' at-the -------~~ 4 4
Crystal River Nuclear Plant. The Crystal River Nuclear Plant is a. Babcock.& Wilcox designed PWR very.similar-to-"- -' - --
the Rancho Seco Plant. In this time he held management :
positions in maintenance, technical support and was the l Assistant Plant-Manager and-Plant-Manager. -- - -- - ~~ ^ ~
i Outage Manager - J. R. Shetler i Mr. Shetler is the Manager, Nuclear Scheduling at Rancho f I
I Seco. He has 15 years of Babcock and Wilcox PWR !
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i 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. ,
Manager, Quality - S. R. Knight i
Mr. Knight has over twenty-five years experience in construction and operation of power plants with initial 3
experience in design, project engineering and test j engineering of U. S. Navy
- nuclear p'ower plants. His recent t
experience has been in design, licensing, construction, test and operation of commercial generating plants ,
including design reviews, safeguards, and waste management,
- - w. - . m m . - .=.a., sQA/QC,cprdrams,forAnventory.,eaterial pand maintenancewAansamrur control.
Manager Nuclear P1. ant - G.. A Coward . . + , -
Mr. Coward has 19 years with the District and has over 16
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years experience with-the-Rancho Seco-Plant: He-har~ held-
positions of Senior Mechanical Engineer, Supervisor Nuclear Maintenance Division and Nuclear Plant Superintendent.
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Manager, Nuclear Projects - J. V. Vinguist V
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 i assigned as Technical Staff Assistant to AGM, Nuclear at SMUD.
1 Manager, Nuclear Training - P. Turner
, Mr. Turner has. worked.ethintthextraining>andinucleemw..ero ma er m
--- -~
- fleids for over-20 years. -In--addition to training--
assignmentt at the Tennessee Valley Authority, and the Insti tute..of. Nucleat Power. Operations., Mr. Turner..was ...+ ~4 - --
Manager of the Nuclear Training Department at Kansas Gas and Electric company.
Manager, Nuclear Engineering - O. G1111spie Mr. Gillispie is an INPO employee on loan to the District.
He has more than 20 years experience in the Nuclear Power fleid and approximately 16 years of commercial nuclear j 4
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experience. Prior to this current assignment, he was Manager of the Technical support Department, and directed the evaluation of technical support activities at nuclear power plants and at the corporate level. He also served approximately 5 years as Manager of INP0's Events Analysis ,
Department.
Manager, Nuclear Licensing - R. Ashley Mr. Ashley has 30 years of experience in atomic energy and nuclear power plant design and operation, with rtsponsibilities 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 m ., m n .- ,rd , o e . g ..,, has par.ticipated.1n..deve. loping: andulmplementing ther Testartu trmwe um programs for-two nuclear- plants-that- received -NRC$ shutdown-~~~*~
orders. He is a registered professional engineer in Nuclear Engineer in.the.. State..of California..-... m+. . .
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1 Figure 2-1 Restart / Inpleinentation i Organization AGM i " NUCLEAR J. E. Ward j .
i MANAGER RESTART / SITE OA l IMPLEtENTATION s. Knight MANAGER l ,
D.C. Poole a l ru:
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- SYSTEM J i
REVIEW l l
& TEST -
MANAGER MANAGER MANAGER T1ANAGER MANAGER ,
PROGRAM PROGRAM OUTAGE NUCLEA_R NUCLEAR NUCLEAR NUCLEAR NUCLEAR DIRECTOR DIRECTOR MANAGER PROJ,EC}S ENGR. PLANT LICENSING TRAINING T. C. J. J. R. J. .; D. G. R. P.
! Lutkehaus Field Shelter VinquisE Gillispie Coward Ashley Turner 5
, F PERFORt1ANCE ItPROV$ TENT mTiOu a
-- Indicates functional accountability
, ; for restart and improvement program
- f. 2-17 responsibilities.
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2.3 ACTION PLAN ACTIVITY MANAGEMENT PROCESS
'N (G '
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 Olstrict has estr.slished the 4
necessary organizational structure, assigned appropriate authority 3
and responsibility and developed the necessary guidelines to assure the program accomplishes its intended near and long term objectives.
The organizational elements instituted to meet these requirements i include:
.1 The establishment of the Recommendation Review and Resolution Board (RRRB-)u w v..m m.
The Board is a nine member multidisciplined group of individuals 7 .- -with nuclear, experience-and -training-drawn-from-SMU&,1 mother ~~~ '- ' ~""
l utility with a B&W NSSS, the NSSS Vendor, and the Plant
- Architect Engineering firm. The functions of this Board are
-(a) to screen'reccmmendations'for clarity ~and duplication,'(b)'
i evaluate issues and recommendations,.and (c) recommend the appropriate disposition and priority for the recommendation based on its technical merits.
To guide the Recommendation, Review, and Resolution Board in
\j making these technical assessments and prioritizing the
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implementation actions, the following guidelines have been established:
Would implementation of the proposed recommendation:
a) Reduce reactor trips i
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 i
. f)- Indicative of programmatic deficiency- - -
g) Sign.ificantly improve. reliability /availabillty . - - ~--
t j If the recommendation meets any of these guidelines and is
. -- determined- to be -valid, -the RRRB provides- inttial prioritization --'-~-- -
In accordance with engineering judgement as follows:
t CRITERIA FOR PRIORITIZING RESTART SCHEDULE ACTIONS TO BE COMPLETED PRIOR TO RESTART (O / OR COMPLETION OF THE RESTART TEST PROGRAM l
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i Assure plant remains in post-trip window Assure compliance with license requirements l
Minimize the need for operator action outside the-control room within the first-10 minutes after an event i
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ACTIONS TO BE' INITIATED AS PROMPTLY AS i
PRACTICABLE, SCHEDULE DEVELOPED, RESOURCES ASSIGNED AND MAINTAINED UNTIL COMPLETED (IT IS THE INTENT TO INITIATE THESE ACTIONS AS S00N AS PRIORITY ONE ACTION COMPLETIONS 4
MAKE RESOURCES AVAILABLE)-
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Enhance ability to remain in post-trip window l automatically i
- Reduce reactor trips Reduce challenges mto.. safety 4ystessemmu w.m- ~we - mee l
l Produce near-term-programmatic--benefits---- - - -
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_ ACTIONS TO BE_ PROGRAMMED JOR.THE. LONGER. TERM - ~~. --- --
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- Improve reliability I
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- Improve availabillty-i Major programmatic enhancements l In addition to screening and validating recommendations, the i
RRRB is also'the group tasked with, developing and maintaining i the master data base which records and tracks each I -
recommendation through its life cycle.
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1 _ . _ _ . , ,__ __. . _ . . . ~ . . - . - . . _ _ . _ _ _ _ _ _ _ _ . . ~ _ . . _ _ . . . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ .
Once the RRR8 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 l
is then sent to the PAG for disposition. The PAG with the approval of the AGM, Nuclear,. determines the course of action for these recommendations.and sends them to the appropriate departments for implementation through the Action Plan or the development of justification as to why the recommendation is invalid..
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.2 The establishment of the Performance Analysis Group -- ~"--~ ~"- -
- 4, _ This group is made.up,of the. Nuclear... Department.. managers-or ---. ~ -- - - -
their designees that report directly to the Assistant General l 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.
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 i I
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d through existing programs or whether adjustments are necessary g 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 reccmmendation. .
4 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 to assure the near and long term.
objectives are met will be developed by the Nuclear. Projects Manager and approved by the Performance Analysis Group and the 1
- Assistant General Manager (Nuclear). This includes changes to the priority of individual action items.
!O .3 Implementation and-Close-out fr. m . e n .e.nrn y - n -m Approved actions are implemented by the appropriate line
- ,,, organization.in.accordanca with the. approved.Quattty 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 l actual implementation documentation. The closecut document is 4
approved by the implementing department manager and forwarded to the Quality Department for final verification and closecut. l i
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Quality ~ Assurance performs a verification of.the implementation' ,
of.the recommendation. This verification. audit can apply.
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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 Performan'ce Improvement Manager for logging and filing.
.4 Action Plan Activity Tracking and Reporting
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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 deficiency it is being implemented to address; This is
- m. . w. my 3 , w . . . pa r t I c u lar 1 y .impor ta n ta s i nce eac h ths sva ror. def.1 et ency1may n $r ihm enWm require -severa1 act!ons to*accomprlish closurerandTpart'icutar' *"~~ ' ~ ~ ~
action may be required as part of the resolution of more than-
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- , , , one issuev .The~ status. of-each< activi-ty>wil>1 be eainta-ined,~~"" ~~ "
monitored, and reported on a weekly basis during program implementaticn prior to Restart and on a monthly basis following Restart Power Ascension -tes' ting.- -----"
The Outage Manager is responsible for maintaining-the schedules for the implementation activities which impact plant hardware and management or programmatic issues. These schedules will 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. l 2-23 l
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2.4 REVIEW MEETINGS AND REPORTS CN U
The AGM-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.
.1 Internal
- On a monthly basis, the AGM-Nuclear will meet with the Rancho s Seco Implementation Committee of the Board of Directors to review in detail, progress of the action plan. At the
.- ., _ subsequent. full. Board.. meeting..he wl.ll-present an overview;ofre a:v.m:>.o ;-am his report.
On a monthly. basis, the. AGM-Nuclear.will . meet withmthe General m ~o . ~+.
- Manager, Assistant General Managers and relevant staff to review in detail progress of the restart and performance improvement
. .. plan. Informally,-the AGM-Nuclear-will-provide the-General -~- - --
Manager with daily updates.
On a monthly basis, the AGM-Nuclear will provide a summary ,
status report for District employees.
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.2 External O
On a monthly basis, for the period beginning August 1986 and i extending to six. months after the beginning of restart heatup, the AGM-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. the period August .1986 and extending to six months after the beginning of' restart heatup, the AGM-Nuclear will meet with the Independent Review Group to review in detail, progress of the Action Plan. These meetings willcontinueonaquarterlybasisforaoneyearjeriod.
On a monthly basis for the period August 1986 and extending to six months after the beginning of restart heatup, the z.. reg AGM-Nuclear w111- send a.. monthly; wrl'ttenoreportroff. Action P:larm armmewoS progress to the following- organizationt:---
American Nuclear Insurers .-.- -
Institute for Nuclear Power Operations B&W Owners Group Executive Committee 1 Supervisors, Sacramento County' -- -
1 Supervisors, Amador County i
Supervisors, San Joaquin County Chairman, California Energy Committee l
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2.5 TRANSITION ACTIONS l The' District's 1990 Plant' Performance objectives are to achieve
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I performance ~1evels which will place Rancho Seco among the top performers.in the. United States. e t .
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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-i termination'of RIO is planned after the restart power ascension ,
testing is complete and sufficient actions have been'taken on the long term performance improvement actions,to assure they can be l managed and implemented by the normal line organizationc A nem::.omm.mrmtrans t tion ,pl an wi l l; be:developednte' assure-thatzthi revoluttore.4 nTm',C tWWm1 L
p _ -.. --- - -the organizational, structure-isederly and til requiredlittions wa ~~'-~~~
anticipated and managed.
I The long term objectives will be achieved through the implementation j j
of the performance improvement items identified in this document and the implementation-of supplemental' programs.-In'particular,"the* ' ' ~~~ ' ~ ~ ~
j District intends to implement the programs necessary to a)-identify-the components critical tc' power production, b) to monitor, trend. -
l and evaluate unavailability contributors, c) to undertake a plant
! specific risk assessment, and'd) to implement those features of the ,
precursor review program necessary to achieve a high quality lessons l 1. earned program.
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.. - - . - - = . . - .. - . . . . - . _ - . - . . _ - _ _ _
i The transition of the special NRC and IRG oversite activities for this Action Plan to those consistent with normal industry practices
! should occur when the-restart actions are complete and those near term actions'to be initiated prior to restart and continued to
- l. completion have been initiated and a plan developed in sufficient .
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i 3.0 O
PERFORMANCE IMPROVEMENTS UNDERWAY PRIOR TO THE 12/26 EVENT.
In late 1984, the Board of Directors of the Sacramento Municipal Utility District recognized that the District had a significant i number of challenges facing them. Foremost among those challenges
- was the need for improvement in the operation of the Rancho Seco Nuclear Generating Station. The Board recognized that the Rancho
- Seco problems had developed over a number of years through the joint attitudes and performan'ce of the Board, the Staff, and plant
~
personnel.
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To a large degree, these failures were made evident by the-overburden the District's staff felt in responding to the large number of '
l changes required to. implement the TMI-2 lessons learned in areas
! incluaing plant modifications, personnel performance, management ,
l systems, analytical capability, training, and organizational ,
i structure. The Board also recognized that the dynamics of the public power arena contributed significantly to the District's arrival at
, its current situation. -The Board was taking corrective actions on these issues when a transient occurred on December 26,.1985 at Rancho Seco which emphasized the need for further action.
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 I plant performance. .The thrust of:the improvement program was to deal iO 3-1 j l
with a large spectrum of management and organizational issues including:
Establishment of a commitment to excellence in performance at Rancho Seco, including strengthening the technical competency of the people and the organization.
The effectiveness of interface activities within upper management and between departments.
Organizational streamlining staff enhancement and other organizational improvements.
Effective attention to detail.
Upgrading of the training organization, training facilities, and training programs Establishment of a clearly defined maintenance program.
Establishment of an effective systematic troubleshooting program i
- Development of a comprehensive root cause analysis program 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 O Spring of 1985, and the program continued through the summer.
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.3-2
4 3.1 PROJECTS UNDERWAY PRIOR TO DECEMBER 26, 1985
.1 Staffing and Organization Significant to the conditions determined in the 1984 study was the recognition c' the sizable growth in plant staff, without a a
corresponding restructuring or expansion of the management staff. A new organizational structure was approved with six Nuclear Departments reporting to the AGM-Nuclear:
i
- 1. Nuclear Operations
- 2. Nuclear Engineering
- 3. Quality s 4. Nuclear Training
, 5. Nuclear Licensing l
l 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, t
1.
Numerous other structural changes occurred within the various departments. The purpose and effect has been to reduce the
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diverse managertal requirements upon individual supervisors and superintendents by establishing new divisions and alignments which 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 is rcceptive a the programmatic approach and committed to attention-to-detail and accountability.
.2 Training Program 7
- 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 elevated to departirental status to ensure top management involvement.
I h
v As of June 1985, the training organization became a department answering directly to the Assistant General 3-4 l
Manager, Nuclear. The position of Training Manager was I
O established and filled from outside the District organization, bringing a new perspective and experience level to the department.
With the recent transfer of Emergency Planning training to the training department, all of the plant training programs are now under the training department with only one exception. This exception is the Fire Brigade training program which will also be tra'nsferred to the Training Department as soon as qualified personnel can be hired.
These management and structural changes, together with the training procedure and policy revisions identified in Section 4.C.3, 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 organization functions.
- 2. INPO Accreditation Effort fhe District has been committed to INPO Accreditation for its training programs for the past several years in an effort to improve the overali training progran. The District is using a phased approach for this effort. To focus on obtaining accreditation for various department functions on a sequential basis, starting with operations.
3-5
f 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 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 i 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 1'
- Maintenance Supervisors and dedicated Preventive Maintenance Crews in the Mechanical, Electrical, and I&C -
Groups.
3-6 i
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- 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.
.4 Quality Assurance Rancho Seco has undergone a number of changes in the quality 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 organization (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 ,
t 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 1 of twenty two qualified inspectors that cover the area of I&C, electrical, mechanical, civil, NDE and concrete structures are now on board as District employees. For outages, the inspectors O
3-7
h-are augmented as needed by contract personnel. The QC O' inspectors are involved in source inspection, receipt inspection, construction and maintenance activities.
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 maintenance operations, both corrective and preventive maintenance. The 4
design engineering review group monitors the design control I
procedures and assures that quality requirements are added to both design specifications and purchase requisitions. The addition of Quality Engineering has increased the capability of the organization and, therefore, it's ability to perform its intended function.
A major revision of the QA Manual was undertaken in 1985. An-extensive effort was made to upgrade the manual and separate it into two sections consisting of (1) 18-point policy section and-(2) the Quality Assurance Procedures. The manual das review d I and approved oy the MSRC and NRC via the 100FR50.54(a) 4 requirement. The policy section will become part of the USAR.
.5 Systeuatic Troubleshooting In March 1985, Rancho Seco was shutdown for a scheduled 90-day refueling for cycle 7. The plant was' returning'to service 94 3-8
days later when a RCS vent line cracked, causing a shutdown due
' to excessive loss of primary coolant. The subsequent investigation and repairs led to a greatly expanded IE Bulletin 79-14 pipe support program, which did not allow restart until late September. Between then and the end-of-the-year three reactor trips occurred. Upon each occurrence, a special systematic troubleshooting program (based upon the D' avis-Besse NUREG 1154 Appendix B methods and criteria) was implemented, j This was in 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.
1 In each case, the implementation of the Systematic Troubleshooting method led to a greater understanding of the event and determination 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.
.6 Root Cause Program l ,
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 finding to the Management Review Team which is made up of O
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the Nuclear Department Managers and AGM-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.
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. 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 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.
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4.0 O
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 4B provides a description of the management, operations, and administrative process improvements for each of the Rancho Seco functional areas as developed by the respective department i
managers.
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- Section 4C provides a description of the plant modifications and maintenance improvements developed by the functional organizations.
Section 40 provides a description of the systems -review and test -
program.
Note that a significant number of the actions consitted to be
[ completed prior to Restart have been accomplished and are awaiting ;
systernatic closure by the QA process.
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.l 4A SYSTEMATIC ASSESSMENT PROGRAM J
J 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 - Stop Trip Program t
Plant Personnel Interviews I
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4A-1 1
O 4A.1 PRECURSOR REVIEW PROGRAM
.1 Objective The objectives of the Precursor Review Program are to systematically review historical documents and recommendations 4 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 preclu'de 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 Review, Recommendation, and Resolution Board (RRRB) for disposition.
.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 STOP-TRIP Program) consists of the following:
(
4A-2.
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All Transient Assessment Program (TAP) Category C f'%
transients will be evaluated and investigated for their appilcability and impact on Rancho Seco.
All Category 8 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.
All recommendations for Category A' TAP transients will be reviewed to determine their applicability to Rancho Seco.
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- All Rancho Seco transients, starting from the Rancho Seco " light bulb" event in 1978, will be reviewed.
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.
i
- 2. Other Documarit Reviews 1
In addition to the review of the TAP data, the following j documents will be reviewed by a multi-discipline !
experienced team:
k O a. Rancho Seco Licensee Event Reports and Occurrence U 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 i
Enforcement
- d. Notices /Ctrculars issued by the NRC Office of Inspection and Enforcement
- e. Babcock and Wilcox Reports (Preliminary Safety-Concerns, Site Instructions, and other relevent BAH reports)
This program provides for a reverse chronological review starting from 1985. Prior to start-up, documents dating back to March 1978 will be reviewed.
. 3 Criteria and Methodology for Precursor Evaluation Each document will be reviewed to determine whether issues are applicable to Rar.cho Seco. For each docucent a Precursor Review Checklist will be completed. For those issues which produce a recommendation, the Precursor Review Recommendation j Form will be completed and forwarded to the RRRB.
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.4 Schedule Evaluations of all recommendations will be completed and prioritized before plant restart.-
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4A.2 DETERMINISTIC FAILURE CONSEQUENCE ANALYSIS
.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.
e
.2 Program Scope 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 tdentify affected systems which challenge or adversely effect the capability to mitigate transient conditions'.
O 4A-6 l
.3 Methodology i Each system will be' reviewed as described below:
- 1. Loss of Electrical Power Teams will analyze each 410V 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" individually. (Note: The
" failure" is an assumption, no physical positioning at i
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 panels off a common 480V bus. Once the consequence of loss of the individual loads is evaluated, the loss of the source (s) 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--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 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.
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I- .5 Schedule 1
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This project, including the evaluation of. recommendations j
- . generated from the review, will be completed and prioritized I i j before plant restart.
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4A.3 O
B&W OWNERS GROUP PROGRAMS - STOP-TRIP PROGRAM
.1 Objective The objective of the B&W Owners' Group (B& HOG) Trip Reduction and Transient Response Improvement Program (STOP-TRIP) is to reduce the number of trips and complex transients on B&W plants and to 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 plant average trip frequency will be less than 2 per year.
, 1 By the end of 1990, the number of complex transients, as
(' 'g classified by measurable parameters (Category C as defined by the B&W Owners Group), will be reduced to less than 0.1 per year per plant based on a moving three-year average.
The District will fully participate in the STOP-TRIP Program.
Participation in the STOP-TRIP and other B&WOG Programs will ensure a broad perspective is taken with respect to plant improvements as well as to allow the other-B&W Owners to benefit from the Rancho Seco Plant Improvement Program.
.2 Program Scoce The STOP-TRIP Program is similar in many ways to the District's Plant Performance Improvement Program. However it is designed 4A-10
__ . - _ - - ~ _ _ _ _ . . _ _ _ _.. _ . _ _ _ . - _
to allow all utillties.with B&W-designed plants to provide input and evaluations. The program is an extension and i
expansion from previous B&W Owners Group activities aimed at reducing the number 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 current District program is completed.
The major activities of the STOP-TRIP. program are:
4
- l. Independent Sensitivity Study
- 2. Risk Assessment Review
- 3. . Operating Experience Review,
- 4. Operator / Maintenance Reviews
- 5. System and Component Reviews The details of the individual projects are in some cases still being reviewed with the NRC staff and are being covered in B&W '
, Owners Grcup submittals to the NRC.
All B&W Owners Group reports-issued since 1979 (the'BWOG was i
formed in response to the 1979 THI-II Event) will be reviewed for recommendations' applicable to Rancho Seco, and to assure l
1
, 4A-ll
that past recommendations have been appropriately considered.
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The recommendations will be evaluated as to their applicability to Rancho Seco as well as whether they have already been.
Implemented or not. Any that are applicable and outstanding will be forwarded to the RRRB for implementation / evaluation prior to restart.
l
.3 Methodology / Procedure The District's STOP-TRIP Program Coordinator (STPC) is responsible for providing the interface between the District and the B&W Owners Group Programs. The STPC shall review the recommendations of these programs and compare them to the O recommendations of the PP&MIP. Recommendations that are not already in the PP&MIP will be submitted to the Recommendation Review and Resolution Board (RRRB) for evaluation and processing through the normal action plan process.
.4 Schedule-Time of Performance In recognition of the fact that the B&W Owners Group STOP-TRIP program is a relatively long-term-project, any items issued by the STOP-TRIP program after startup will be addressed through a leng-term program for similar treatment. Several of the SWOG~
programs and projects are on longer schedules than can be accepted by Rancho Seco. ,As such, the District will procede 4A-12
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l with these to the degree that the Rancho L Seco requirements I dictate. Results will subsequently be shared' with the other L.
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, -BWOG participants. .
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f' 4A.4 PLANT INTERVIEWS
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.1 Purpose ,
The interview program is to surface previously unresolved, but I
"known", problems which can (1) cause reactor trips and/or contribute to the severity of transients, and (2) degrade plant i
reliability or the optimal performance of the operating personnel.
l .
.2 Project Scope I .
This program will interview personnel from key plant functional
- groups. These persons will be encouraged to identify systems,
- components, or operational problems and concerns of which'they are aware and provide recommendations on how to resolve them.
Volunteers will be requested, and all volunteers will be interviewed. A minimum number of interviews have been established within each functional group and the interview coordinator will select personnel for interview if there are insufficient volunteers.
.3 Interview Methodology The interview program is-to cover a cross section of plant I personnel and is intended to encourage personnel to identify issues or concerns of which they are. aware that, when resolved, i
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4A-14 I
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can contribute to optimal, reliable, or improved operation.
Os Each interview will be documented using an interview form.
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.
i The Interviek Project Coordinator consolidates the compiled list of concerns / recommendations. He forwards the recommendations to the RRR8 using the Recommendation / Resolution O sheets. The Interview Program Coordinator assures that the concerns and recommendations have been acted upon and
. dispositioned to the RRRB.
.4 Schedule The program interviews, including evaluation and disposition of the recommendations will be completed prior to plant restart. l Approximately 180 volunteers were interviewed. Some 1600 recommendations were developed prior to consolidation to eliminate duplication. The RRR8 is now acting on these recommendations.
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4A-15 l
4B MANAGEMENT, OPERATIONS, AND ADMINISTRATIVE PROCESS IMPROVEMENT t
This section identifies the program enhancements developed by the District's department mankgers which are being implemented to address the deficiencies contributing to the performance record and the ,
j i
i December 26, 1985 event. These programmatic actions form the broad i
framework for the implementation of the findings from the systematic review process. Each programmatic action area is discussed in i general, followed by specific commitments which address the deficiencies related to the programmatic area.
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i 48-1 i
4B.1 MANAGEMENT EFFECTIVENESS l.
Management and management process effectiveness have a major impact 4 on the ability to operate the Rancho Seco Nuclear Generating Station :
in a safe and reliable manner. .
1 The objectives of these actions are as follows:
1 1
- a. To develop guidelines and agreements by which the SMUD Board, e
as the governing entity, can improve its effectiveness in directing and monitoring the District's activities and obilgations 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 Assistant General Manager-Nuclear, General Manager and Board.
- c. To enhance the management process in support of the safe and-1 reliable operation of the Rancho Seco Nuclear Generating-Station.
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program.
- O 48-2
- 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.
.2 Actions to be Initiated as Promptly as Practicable. Schedule Developed, Resources Assigned and Maintained Until Completed
- 1. Board of Directors / General Manager Imorovement
- a. Establish within the Board, guidelines and agreements by which the Board, as an entity, can more ,
effectively set policy and direction,
- b. Establish written performance measurement criteria, and a performance review process, for the General Manager (GM).
- c. Clarify the Board / General Manager working relationship-in writing, including the reporting desired by the Board from the General Manager.
- 2. Corporate Management Improvement
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- a. Assess current corporate-support interfaces with' the I b Nuclear Organization and make recommendations to the I
Assistant General Manager-Nuclear and the General' Manager regarding improved management of interorganizational working relationships. .
- 3. Nuclear Program Management Improvement i
3 a. Develop and implement a Rancho Seco Business Plan for use by the Board of Directors. I
- b. Establish a comprehensive, cohesive and clearly understandable set of GM and AGM-Nuclear policies'and i
i practices which provide upper tier direction for
! similar efforts at the functional manager and supervisory levels.
- 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.
- d. Upgrade management programs and practices in the areas of functional planning, decision making, problem solving and interdepartmental collaboration.
48-4 l
- 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.
- 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.
- g. Develop a program for improving communications skills of Nuclear Department managers in presentations to the Board of Directors, the public, and. staff.
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- h. Establish a department Human Resource Management I program which includes:
- 1) identification of priority management development / training needs and the appropriate means for addressing each; 48-5
- 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.
- 1. Improve Department media and community relations by establishing a more proactive media / community outreach program.
j . Improve Nuclear Department interfaces with all other I
Departments in the District by instituting additional interdepartmental communication and problem-solving processes on a regular basis.
- k. Develop a Rancho Seco Facilities Master Plan.
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48.2 QUALITY AND QUALITY ASSURANCE An effective quality program at Rancho Seco is important to achieve the near and long-term performance standards and objectives of the )
01 strict. .
]
The objectives of these actions are to improve the overall effectiveness of the Quality Assurance effort and to assure the benefits of Quality Assurance are realized in the near and long-term.
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program
- 1. Reorganize the Quality function at Rancho Seco to enhance the Site Quality Assurance Department, providing increased focus in the following areas:
- a. Quality Engineering
- b. Quality Control
- c. Surveillance
- d. Vendor Qualification and Source Inspection
- e. Nuclear Program Audits O .
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- 2. Develop and implement the procedures and processes necessary to independently verify the effective closure of the actions identified for the Action Plan. I l
- 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.
- 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 s and implemented.
- 5. Increase the Eite QA Department staff to assure the added demands of the Action Plan and changes in responsibilities
- can be effectively implemented.
.2 Actions to be Initiated as Promptly as Practicable, Schedule ,
l Developed, Resources Assigned and Maintained Until Completed i
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- 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.
48-8 -l l
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- 2. Identify and develop enhancements to the QA program to address any programmatic and management areas that have associated de'ftciencies from a quality perspective.
.3 Actions to be Programmed for the Longer Term ,
- 1. Develop and implement the necessary policies and
- procedures to establish a more proactive quality program,
.i
- which will also improve the effectiveness of audits.
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! 48-9
48.3 TRAINING In addition to subdividing training improvement actions by priority, it is also useful to subdivide into improvement areas. The actions presented here are divided into: management controls, facilities and .
resources; and specific instructional (Operations, Emergency Planning, Other).
.1 Management Controls, Facilities and Resources These improvement actions are those required to bring the District's Nuclear Training Programs up to a state-of-the-industry condition. The intent of this action is to provide long term results and no items have been identified requiring completion prior to restart.
- 1. Actions to be Initiated as Promptly as Practicable, 2
Schedule Developed, Resources Assigned and Maintained Until Completed i
- a. Continue the upgrade of Non-Licensed Operator Training to maintain INPO Accreditation. i
- b. Initiate the process of achieving INP0 Accreditation
! for the maintenance train ng area.
l l
48-10 I
- c. Develop the plan for installation of a computerized
( Training Information Management System.
\s,
- d. Develop plans for centralized and secure storage of training records. .
I
- e. Develop or purchase a Rancho Seco Simulator Baseline Data Information and Tracking System consistent with l the Simulator now being purchased.
4
- f. Incorporate short term training items (lessons learned) into permanent training materials.
t
- 2. Actions to be Programmed for the Longer Term
\
- a. Complete the purchase and installation of a' plant specific simulator.
- b. Complete the staffing of the Training Department with SMUD employees.
1
- c. Complete and maintain INPO accreditation for the
. remainder of the Training Programs.
i
.2 Specific Instruction Actions Items - Operations 4
48-11
The specific instruction action items for Operations are
\
.Q( , detailed below:
- 1. Actions to be Completed Prior to Restart or Completion of the Restart Test Program .
- a. Train licensed operators on Emergency Operating Procedures, including the changes resulting frcm the December 26, 1985 event, those revisions resulting i from the EOP/ATOG review, and all completed recent 4 design modifications.
- b. Train licensed operators on the loss of ICS/NNI, including those p.rocedures added or revised as a result of the December 26, 1985 event and related a
design modifications.
- 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 4
wrenches. Incorporate into requalification training.
- d. Train operators on the specific lessons learned from the December 26, 1985 event. These include such items as the makeup pump failure, overfilling the makeup tank, cooldown rates, reactor vessel head bubble formation, and the functioning of valve i
O actuator controls.
48-12
- e. Train operators on watch standing principles, t
(' _,/ including command and control training for shift supervisors, role and function of STA, equipment monitoring.
- f. Retrain operators on health physics requirements associated with their job responsibilities.
- g. Train operators on their job related functions associated with startup testing.
3 t
) .3 Specific Instruction Action Items - Health Physics The specific instruction actions for Health Physics are summarized below:
- 1. Actions to be Completed Prior to Restart or Completion of the Restart Test Program i
- a. Train Health Physics Technicians and Operators on the
- procedure (s) for entry 1.ito areas of unknown radiological conditions.
- b. Train Health Physics Technicians and Operators on proper response to radiological emergencies.
I 48-13 i
- c. Train Health Physics Technicians and Operators on evaluation of radiological effluent discharges.
. 4 Specific Instruction Action Items - Emergency Planning The following specific action items are to improve emergency.
response.
4 1. Actions to be Completed Prior to Restart or Completion of the Restart Test Program i
1
- a. Train assigned maintenance personnel on the maintenance of the Interim Data Aquisition and Display System (IDADS).
) b. Train operators on the operational use of IDADS.
i I
- c. Update Emergency Preparedness Training Instructor Guides, Student Guides, and visual aids to support the October-1986 Drill.
- d. Train assigned personnel on the revised Emergency Plan Procedures.
- e. Provide management guidance to the operating crews (through training) on prioritizing multi-casualty events.
48-14
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! 2. Actions to be Programmed for the Longer Term j i
- a. Develop and implement an improved process for l continuing Emergency Response Organization training. I l
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48.4 OPERATIONS AND OPERATING PROCEDURES-Improvement in operations is brought about by improving management controls such as procedures, organization, policies, and through t
improved training of personnel. Improvements in training are - <
addressed in Section 48.3. The actions to improve management controls are addressed here.
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program
! 1. Issue management guidance, via procedures, defining the policy on procedural compliance. This procedure will ,
provide a direction on what constitutes " procedural
- compliance"'and " procedural guidance".
4
! 2. Correct specific procedural deficiencies identified during the review of the December 26, 1985 transient .
l 3. Review and upgrade the CR/TSC HVAC operating procedures.
l
- 4. Verify technical correctness of E0P changes made since May 1
1985.
- 5. Compare E0Ps to ATOG Technical Basis. -Incorporate identified improvements into E0Ps.
48-16 i
'._.___=...__._....__...__.__...__..___
. _ . . - - = .
- 6. Make the necessary modifications to the design change process to assure that design changes are incorporated into all operating procedures in a timely manner.
4 1 7. Assure operating procedures address the recommended topics .
of Regulatory Guide 1.33 Sections listed below. Implement
- ~ procedures which may be required.
i i a. Section 3 Procedures for Startup, Operation, and Shutdown of Safety Related Power System.
- b. Section 6 Procedures for Combating Emergencies and Other Significant Events.
- 8. Perform a valve walkdown to verify the consistency of i
as-built conditions, P& ids, procedural lineups, and I
component identification of important secondary mechanical systems. Initiate corrective action for inconsistencies which are identified. This compliments the long term Configuration Management actions defined in Section 48.12. Systems included are:
1
- a. Air Ejector / Gland Seal
- c. Auxiliary Steam
- d. Component Cooling Water f
{
- e. Instrument Air j l ,
i i
4 48-17 i
1
4
- f. Main Circulating Water
- g. Main Condensate
- h. Main Feedwater
- 1. Nitrogen Gas
- j. Plant Cooling Water -
- 1. Turbine Electro Hydraulic Control
- m. Turbine Lube Oil l .2 Actions to be Initiated as Promptly as Practicable, Schedule j Developed, Resources Assigned and Maintained Until Completed i
- l. Develop a revised Nuclear Operations organization and i begin staffing at the management level.
- 2. Develop a staffing plan and schedule to meet the needs of i the revised Nuclear Operations organization. This will I
include the needs for.llcensed operators identified as rotational / transfer assignments.
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-l 48-18 1 __ _ _ . -
48.5 MAINTENANCE PROGRAMS AND PROCEDURES
' i V
The quality of the maintenance programs has a direct impact on the l material condition and reliability of systems and components throughout the entire plant. The actions described below are .
Intended to provide District Management with assurance that the material condition of Rancho Seco's safety systems, and those systems required for normal control as well as post-trip control, are such that safe operation may be resumed.
.1 Actions to be Completed Prior to Restart or Completion of the i Restart Test Program '
- l. Inventory Calibrated Test Equipment (CTE) and calibrate .
V and/or control use to prevent use of uncalibrated CTE.
- 2. Identify and assure current calibration of all in-plant instrumentation used in the performance of surveillance testing.
- 3. Rework the makeup pump and return to service.
i
- 4. Complete the in-progress battery replacements (A, B, C, D, E, F).
- 5. Perform refueling interval surveillance of snubbers.
. O 48-19
- 6. Complete rework of terminations in the Bailey Cabinets in
\j the Control Room (NNI/SFAS/RPS/ICS).
- 7. Perform biennial Diesel Generator Inspection and replace turbochargers. ..
- 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 1 and steam generator heat removal applications (143), plant instrumentation required for surveillances, safety related HVAC and the Control Room normal HVAC system.
- 9. Complete Preventive Maintenance (PMs) on selected manual valves.
i
.2 Actions to be Initiated as Promptly as Practicable Schedule j Developed, Resources Assigned and Maintained Until Completed I
i
- 1. Develop a departmental procedure hierarchy and writer's i
guide for Maintenance Procedures.
- 2. Identify and prioritize procedures for generation and/or ;
revision.
i 4B-20 I
- ,, 3. Achieve authorized staffing levels within the PM
, k organizations and activities.
- 4. Develop and/or revise the required programmatic procedures
, for.the PM program to: assign responsibilities, authority .
and accountabilities for the program; establish criteria i
! and define the scope of the program; and define the interface with other work control processes.
i
- 5. Review existing PM tasks and frequency for critical equipment. Revise and augment as required by programmatic 1
selection criteria. ;
- 6. Perform Laboratory Failure Analysis of the ICS SI/S2 -
switches and ICS Power Supply Monitor which were installed
{
on December 26, 1985.
l i
1 4 48-21
__.._-___-__._._.1 1
s 48.6 HEALTH PHYSICS AND RADIOLOGICAL CONTROLS Improvements in Health Physics or radiological controls is brought' '
l about by improvements in organization, procedural controls, and training. Training is addressed in Section 48.3. Organizational and .
?
procedural actions include:
.1 Actions to be Completed Prior to Restart or Completion of the 1
' Restart Test Program 4
- 1. Relieve operators of special HP duties.
1
- 2. Prepare a procedure, for Health Physics Technicians to use i
for entry into unknown radiological conditions.
- 3. Revise setpoints for plant gaseous-effluent monitors to- -
j ensure unambiguous indications.
I.
- 4. Issue a Radiological Event Directions Manual to provide more guidance for abnorrimi situations.
i
- l j 5. Issue new manuals to separate event and instrument procedures from the Radiation Control Manual.
k 2
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48-22
.. . . _ , . _ _ _ _ . . . _ . _ , . . . _ . . . .~ _ _ . , _ . . _ _ _ _ . . _ _ _ _ _ _ _ _ . . _ . - , _ . _ . _ _ . _ _ _ , . . _ . . . . _ . _ . _ .
48.7 10CFR50 APPENDIX I DISCHARGE GUIDELINES O
O The 10CFR50 Appendix 1I Guidelines for radioactive cesium can be exceeded under certain circumstances when the current lower limits of detection are used and applied in conjunction with technical .
specification requirements.
The objective of these actions is to upgrade the analysis and controls to provide confidence that discharges and the cumulative impact of discharges will satisfy the objectives of the environmental discharge requirements.
j . 1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program l
l .
- 1. Evaluate the current radioactive waste analysis methods -- -
- and sensitivity relative to their ability to support l operation needs and the performance objectives of this task.
- 2. Develop and implement the changes in Radiochemistry methods and controls necessary to achieve the objectives of this task.
i i 3. Review and revise the off-site discharge calculation manual to incorporate these changes.
48-23 4
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. 2 Actions to be Initiated as Promptly as Practicable, Schedule-Developed, Resources Assigned and Maintained Until Completed i
i
- 1. Evaluate the design of plant systems with the intent to
- improve the ability to operate within Appendix I .
! Guidelines when operating with primary to secondary leakage. Implement plant improvements as appropriate.
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r 48.8 PERFORMANCE IMPROVEMENT AREA: EMERGENCY PREPAREDNESS An effective Emergency Preparedness program is essential to the assurance of the health and safety of the public should an environmentally impacting event occur at Rancho Seco. Several ,
weaknesses were apparent in the Emergency Preparedness Plan, and its implementation during the December 26, 1985 event.
The objective of these actions is to upgrade the Emergency Prepare'dness program to assure that it is efficient and effective.
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program
- 1. Meet with NRC (Region V) to review / critique this Action Plan.
- 2. Update Emergency Plan a'nd Implementing Procedures to address December 26 event Lessons Learned.
- 3. Establish independent meteorological assessment capability.
- 4. Integrate EP commitments into commitment tracking program, i
l l 5. Evaluate notification / communication system and implement upgrades related to December 26 event Lessons Learned.
t 48-25 i
[ .
- 6. Simplify Control Room dose calculation procedure.
- 7. Implement new PASS procedures including core damage assessment.
- 8. Initiate " mini-drills" program.
- 9. Coordination and support of the Training Group per 48.3
- including
i
- a. ERO identification
- b. Plan / procedure update information
- c. Facilities identification
- j d. Instruction materials upgrde
- e. Scheduling i f. Tracking and documentation j g. Data base management
- h. Management support of ERO participation
.2 Actions to be Initiated as Promptly as Practicable. Schedule Developed. Resources Assigned and Maintained Until Completed 4
- 1. Establish separate onsite and corporate plans.
- 2. Consolidate / cross index EP and Central files.
- 3. Complete multi-parametric data base including EP records
- and schedules.
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48-26 I
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Provide positional analyses for ERO versus District staff.
4.
~5. Define / implement public education program enhancements for emergency response.
- 6. Integrate Emergency Plan Implementing Procedures and plant operating procedures.
I
- 7. Complete installation of notification / communication system including verification, training and drills program.
i
- 8. Redefine Emergency Preparedness maintenance program.
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, 48.9 HUMAN FACTORS Human Factors reviews of the transients in October and December, 1985 i have identified the need for plant modifications, and procedure changes. .
The objective of these actions is-to correct the identified deficiencies and enhance human interface with the plant.
. 1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program
- 1. Implement modifications and procedure changes resulting from the post trip Human Factors reviews:
l a. Provide operator training associated with AFW valves FV-20527 and FV-20528.
i
- c. Improve interface between Security and Control Room I
personnel.
'd. Install long cord on red phone, i
j e. Relabel ICS power supply breakers 51/52.
48-28
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--..,vw .,ry-. ,--_ ,,,.._w v.,..v-.,_,%., ,,-_,,..,y , o- 3..,,
.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained until Completed
- 1. Modify Control Room access doors such that one door is
. -used for exit and one for entrance. .
l l
- 2. Modify red phone power supply to eliminate spurious.
ringing following power supply transfers.
- 3. Assess ~ capability to accelerate implementation of CRDR-modifications (Mod 142) currently scheduled for Cycle 8 If possible accelerate ind cycle 9 refueling outages.
- implementation.
.3 Actions to be Programmed for the Longer Term t a 1
- 1. Participate in the INPO Human Performance Evaluation j
program.
- 2. Implement CROR modifications (Mod 142). These modifications were identified in the Olstrict's submittal
- to the NRC in December, 1985, which documented the results j of the Rancho Seco control. room design review.
t
- 48-29
48.10' MANAGEMENT INFORMATION SYSTEM O
V Plant Records and Database information are very important to the efficient and effective conduct.of support activities for plant design, design modifications, operation, and maintenance. The ..
records and data provide historical information on all the various activities at the site and must be effectively managed to support improvement programs and ongoing programs.
- The objective of these activities is to enhance the records and data management activities to effectively and efficiently support the Nuclear Department's needs.
'e
,,1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program
- 1. Review implementation of NRC Generic Letter 83-28 commitments and develop plan for program enhancement.
- 2. Provide site information systems support for implementation and records management activities needed for restart.
.2 Actions to be Initiated as Promptly as practicable, Schedule Developed. Resources Assigned and Maintained Until Completed
\
48-30 k
. , . , . - - m. . _ . - , - . - , , . , , .,,.,_n _ . , , , -. a . , - - + , . . , , . , - . , . _ . , , - - - , _ .
- 1. Prepare a Data Systems Directory of existing systems with .
data source information.
1 i
! 2. Implement Vendor Data Program-enhancements identified to 1
l achieve the program objectives. .
.3 Actions to be Programmed for the Longer Term I '
l 1. Establish program and policy for development of integrated i
data management system design and implementation.
i
- a. Complete Nuclear Information Management System (NIMS) evaluation.
i
- b. Establish on-site facilities and organization to support NIMS hardware / software.
i i
- c. Implement NIMS program / data.
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48-31
48.11 COMMITMENT MANAGEMENT U Commitment management is an important aspect of the Olstrict's interface with "outside" agencies as well as for the management of the Olstrict's day-to-day activities. .
Improvement in commitment management is aided by improvements in management controls, such as implementing procedures and tracking commitment systems. The actions to improve the commitment management programs are identified below.
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program
! 1. Revise commitment management procedure to include tracking and compilance features.
- 2. Install a new commitment tracking system per .1.1 above.
- 3. Develop system / user documentation for the new commitment tracking system.
- 4. Verify the commitment tracking system database with respect to current known commitments.
- 5. Verify all commitments required prior to restart are
- complete.
, 4B-32
4 r
i
.2 Actions'to be Initiated as Promptly as Practicable Schedule t Developed, Resources Assigned and Maintained Untti Completed a
f
- 1. Verify the commitment tracking system historical I
i
' commitment database. ,
e i
. 2. Each Nuclear Department will establish specific milestones j for reducing its backlog of open commitments. l l
.3 Actions to be Programmed for the Longer Term j 1 ;
- 1. Integrate the commitment tracking system with the Nuclear 4
lv Information Management System. ,
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d 48.12 CONFIGURATION MANAGEMENT Configuration management ~1dentifies, controls, status and verifles the plant document and hardware configurations.
This is an important program which is necessary to assure actions on the part of the various organizations responsible for interacting with the plant do so based on a consistent set of documents which are also consistent with the plant hardware configuration.
1 The objective of these actions is to take the steps necessary to improve the effectiveness of the configuration management system at Rancho Seco.
.1 Actions to be Completed Prior to Restart or Completion of the i Restart Test Program
- 1. Verify that control room drawings are current, in accordance with existing procedures.
- 2. Provide Nuclear Engineering support to plant operations to address and expedite configuration management issues.
.2 Actions to be Initiated as Promptly as Practicable. Schedule
- Developed, Resources Assigned and Maintained until Completed 4B-34 e - - --- , . - - ,e ,.- , - + , - , .w - --+ - - - , - -
. . .. .. .- -. .-. = -- - . ._ . - .. .
i
- 1. Review and evaluate all temporary modifications and close out all existing abnormal tags that need to be converted to permanent plant modifications.
. 2. Reduce the backlog of DCNs. ,
- 3. Develop the System Design Basis documents for NEP manuals' on BOP related systems. !
4
)j . 3 Actions to be Programmed for the Longer Term
- 1. Establish management direct. ion for a Configuration i Management program for Rancho Seco consisting of:
- a. Policy
- b. Specifications
.i c. Computer hardware and software i d. Implementing procedures i
- e. Training
- 2. Establish or upgrade existing equipment and supporting '
documentation identification systems needed for total l ,
plant configuration control.
- 3. Upgrade change control packages that control modification I
from change request through close out. This includes i identification of all affected documentation such as procedures. training plans, and simulator upgrades.
48-35
- - , .-- -a -
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, 4. Reorganize drafting into a design / drafting organization.
\
- 5. Train Nuclear Engineers to utilize these designers to reduce the engineering work load.
~
i
- 6. Develop new, or simplify existing procedures to clearly
- define the review process for drawings.
j
- 7. Conduct a cost / schedule review to determine whether a CAD 3
system can be justified for Rancho Seco.
- 8. Develop or upgrade existing systems to provide verification that configuration documentation reflects the true hardware configuration.
- 9. Develop or upgrade existing systems to provide the status of all documentation and equipment in a timely and accurate manner.
- 10. Develop a work package system for all facility changes.
!O
! 48-36
. 48.13 MATERIALS MANAGEMENT O
The Materials Management program is a key element which supports the achievement of reliable and safe plant operations. This program assures that correct parts are available for the repair of plant ,
j components and obtains and manages the materials necessary to support plant modifications.
The objective of these actions is to enhance the materials management program at Rancho Seco to assure the long-term plant performance goals can be achieved.
.1 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed
\
- 1. Conduct a review of the current Materials Management program.
- 2. Develop and implement an action plan to improve the performance of the Materials Management program.
n v
4C PLANT MODIFICATIONS AND MAINTENANCE IMPROVEMENTS E
This section identifies-specific plant modifications, and maintenance improvements, which are being implemented to resolve the j lessons-learned, recommendations, and programmatic deficiencies .
Identified by the Restart and Performance Improvement Program. Each l subject area is discussed in general, followed by specific commitments relative to that issue.
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4C.1 INTEGRATED CONTROL SYSTEM (ICS) AND INTERFACING SYSTEMS The Integrated Control System is an important element in the control I
of the plant during normal operations. Several changes to the ICS and interfacing systems have been identified by the District and the .
B&W Owners Group which will improve the operators ability to maintain the plant within the post trip window and reduce the potential for reactor trip as a result of failures of the ICS. The general programmatic actions and specifi' design modifications to upgrade the Rancho Seco ICS and associated systems are described in this section.
1.a General Programmatic Actions Changes to the ICS, which offer potential to improve the reliability of the ICS, improve the operators ability to maintain the plant within the-post trip window,-and reduce-the *~ - -"
potential for reactor trips are being developed or have been developed by various organizations.
The objective of this general programmatic action is to 1,mplement those changes which are judged to enhance safety or operability.
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program 4C-2
s s
1 The following specific modifications /or ?ctions will l be implemented, as a minimum, to achieve the j l
objectives noted- I l
- a. Improve ICS Reliability .
t
- Improve reliability of ICS power supplies Implement B&W Upgrades Modify ICS MFWP runback rate / level Substitute RC flow Remove SU FW flow (if feasible)
- b. Improve Status Information Upgrade annunciation to remove ambiguities Provide computer inputs for ABT status Add open/close status lights of ADVs, TBVs,
+
MFW valves Tag instruments with ICS processed signals 4C-3 l
Provide first-out MFP trip monitor
- c. Improve Plant operability by assuring that plant will go to known safe state on loss of ICS
- Review procedures for adequacy in event of ICS failure l
- Review / adjust minimum HFP speed 1 - Trip ICS power on loss of NNI-X, Y, or Z power
- Trip MFP's on loss of .ICS control .
1 i
- Provide independent control-grade backup-~ '<-~~*4
"" 7 AFW automatic level control
- d. Improve ICS restoration procedures l
l - Review procedures for adequacy, correct' as - .l necessary Evaluate (and correct as necessary)-final
> control element position on ICS power
- restore 4C-4
- 5. Inspect electrical terminations within ICS.
.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources-Assigned and Maintained Until Comoleted .
- 1. Develop and implement those design changes or enhancements that were not required to be implemented prior to restart, consistent with their assigned l priority.
4
.3 Actions to be Programmed for the Longer Term
- 1. Actively participate in the B&W Owners Group efforts to upgrade or replace the Integrated Control System.
l.b Actions to Reduce the Impact of Power loss on ADV's and TBV's The loss of power to the ICS during the December 26, 1985 event resulted in the ADV's and TSV's falling to mid-stroke due to the bi-polar nature of the ICS. This resulted in an overcooling transient following the reactor trip.
The objective of these actions is to implement plant features which are effective in addressing this deficiency. To accomplish this objective, the following actions will be taken.
O 4C-5 i
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program
- 1. Provide controls in the Control Room from which the operator can operate the ADV's and TBV's, and which ,
will cause these valves to remain closed on loss of ICS (DC) power.
l.c Actions to Address the Adequacy of the ICS Power Monitors The DC power supply monitor for the integrated control system was identified following the December 26, 1985 event as a single failure point which could contribute to the loss of ICS.
The objective of these actions is to evaluate the potential ,
u, .- , _m . -... _ _ .. .!mpact.of .the. power supply -moni tore and 2 to impiementachanges ~a~~ +->~~e which will effectively improve the reliability of the power supply. -
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program I
i
- 1. Determine the contribution the ICS Power Supply
, t nitor had in December 26,.1985 transient.
4 I'
- 2. Evaluate the potential improvement to ICS reliability if redundant PSM's are installed.
4C-6 l
- 3. Determine the potential benefits to be obtained through the installation of independent PSM's.
- 4. Implement design improvements or document justification for not implementing PSM modifications. ,
l.d Actions to Imorove the Status Indication on loss of ICS Power The status indicator window logic for DC power status to the ICS was not adequate to provide the operators with sufficiently clear indication of the ICS DC power status to provide for prompt operator recognition of the cause of tiie loss of ICS.
The objective of these actions is to implement design changes which are effective in addressing this deficiency. To im-, -, m, . ,a. .. . ..accompiishs.this objective.,the: fol_ lowing- actionsuwi%< bet-tqukerte m--e
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program
- 1. Provide separate windows for the status indication of the following:
l 4 - ICS Trouble (fan failure, power supply failure). l 1
ICS Failure (Loss of DC Buss).
O 4C-7
s i.e Actions to Evaluate the Failure Consequences of Various ICS Inputs, Outputs, and Components a
The ICS has undergone a number of changes since a generic failure / consequence analysis was performed in (approximately) .
1980, BAW1564. These changes include not only physical but i philosophical design basis changes.
I i,
The objective of these actions is to update and evaluate the impact the changes have had to the overall functional
, performance of the ICS relative to individual failures of various ICS inputs, outputs, and components. To accomplish this objective, the following actions will be taken.
.1 Actions to be Initiated as Promptly as Practicable,
- e. .m - . , .
Schedu l e Deve loped ede source s 6Arttuned 5 and r Me t n ta ked* uwa-w'w= " : '
1 Untti Completed Participate in B&W Owners Group efforts to perform a generic failure / consequence evaluation of the Model l 820 ICS. If the BWOG FMEA is.not initiated by August 1, 1986 perform an independent, Rancho Seco specific FMEA.
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- .2 Actions to be Programmed for the Longer Term i l 4C-8 l
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- Evaluate results and applicability of B&W Owners Group evaluatten findings and recommendations to Rancho Seco.
- Conduct supplemental evaluations as required to achieve Rancho Seco' specific information. .
- Develop and implement applicable modifications to Rancho Seco ICS.
1.f Actions to Address the Adequacy of ICS Power Supply The December 26, 1985 transient was initiated when the ICS DC i power was interrupted by action of the single Power Supply Monitor. Operator action 26 minutes later reclosed the
~ breakers providing power to the DC Power Supplies. During the
- power outage, the ICS Hand / Auto. stations were inoperable and
% . y. ~ . ,..- . r - de v i c e s control l e d e byr the:ICS :re s ponded* Tswhen 7t ctri v i ngra= *"'w'3 """
zero volt DC control signal. ,
1 The objective of these actions is to evaluate, develop, and implement beneficial design changes to reduce the likelihood that this event would reoccur. To accomplish this objective, l the following actions will be taken.
.1 Actions to be Programmed for the Longer Term
- Evaluate need for DC Bus battery backup based on
, reliability of power supplies, recent modifications, etc.
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- Evaluate Hand / Auto station backup power.
- Participate with B&W Owners Group to enhance ICS and .
related power supplies.
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- Develop and implement appl'icable modifications. ,
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l 4C.2 NON-NUCLEAR INSTRUMENTATION (NNI)
D b
The Non-Nuclear Instrumentation is an important element in the control of the plant during normal operating evolutions. Several changes to the NNI have been-identified by the District and the B&W .
Owners Group which improve the operators ability to monitor and control plant parameters. The general programmatic and specific actions to upgrade the NNI are described in this section.
2.a General Programmatic Actions Changes to NNI, which offer the potential to improve the reliability of the NNI, and improve the operator's ability to s minimize reactor trips and maintain the plant within the post trip window, are being developed or have been recommended by various organizations.~= ...t.-
The objective of this general programmatic action is to assemble, review, and implement those changes which are judged to be safety or operationally beneficial. To accomplish this objective, the following actions will be taken:
.1 Actions to be Initiated as Promptly as Practicabih Schedule Developed, Resources Assigned and Maintained Until Comoleted 4C-11 l __ __. _
- 1. The following specific morilfications/or actions will
,\ !
q) be implemented, as a minimum, to achieve the
- objectives noted
- a. Improve NNI .
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- Improve reliability of NNI power supplies i
- Upgrade NNI Modules to latest B&W models.
- Update NNI drawings to incorporate discrepancies identified in the deterministic failure analysis and the District's o.1 going efforts to upgrade plant performance.
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- b. NNI Status Information l - Upgrade annunciation 1
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- Provide computer inputs for ABT status l
Tag instruments with NNI instruments l
Provide adequate NNI parameter trending in IDADS (including recorder replacement) 4C-12
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- c. Improve Plant operability by assuring that plant will go to known safe state on loss of NNI
< - Provideg automatic trip of ICS power on loss of NNI-X, Y, or Z power .
- Revise procedures for loss of NNI
- Review NNI restoration procedures for adequacy in event of-NNI failure
- 5. Inspect electrical terminations within NNI.
.2 Actions to be Programmed for the Longer Term
- -, ..c-- - _ .1. Actively par.ticipate41nz the 8&Whers-Groupiefforts==Mm ~~
to upgrade or replace-the Non-Nuclear Instrumentation equipment.
2.b Actions to Address the Adequacy of NNI Power Monitors As a result of the December 26, 1985 event, the ICS power supply monitor (PSM) was identified as a single point which
, could cause a loss _of ICS DC power. The NNI has much the same power monitoring configuration as the ICS. The objective of i
these actions is to evaluate the potential impact of these
, power supply monitors and to implement changes which will effectively improve the reliability of the power supplies.
4C-13 1
- - - - - . - - - . .~ -. .. . , , - , - , _ . . . . _ _ . .
.1 Actions to be Completed Prior to Restart or Completion of (O,) the Restart Test Program
'1. Evaluate the potential improvement to NNI reliability if redundant PSM's are installed. .
- 2. Determine the potential benefits to be obtained through the installation of independent PSM's.
- 3. Implement design improvements or document justification for implementing PSM modification.
2.c Actions to Address the Adequacy of Status Indication for NNI, and Affected Instrumentation on Loss of NNI Power 7,. 2,.c, ,-~, . r Anal y s i s n ( DFA) : h as a de te rmi nedr th a tz ar l os s roCNNI;X;r-Ylrorm Za m N > " "
power will result in the loss of critical signals to the ICS and cause control room indication to become ambiguous. This situation results in a reactor trip and loss of indication necessary to maneuver the plant. The following modifications together mitigate the consequence and reduce the dependence on operator actions.
.1- Actions to be Completed Prior to Restart or Completion of the Restart Test Program s
.n 4C-14'
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- 1. Tag NNI-affected indicator / recorder l 2. Trip ICS power on loss of NNI-X, -Y, or -Z power I.
- 3. Make modifications to ICS to provide automatic .
control of ICS/NNI power.
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. i l 4. Provide separate annunciator windows to indicate:
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j Loss of NNI-X (DC)
Loss'of NNI-Y (DC)
- Loss of NNI-Z (DC, switching supply)
I - NNI trouble (fan failure, single power supply.
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4 4C.3 FEEDWATER AND STEAM SYSTEMS a'
The design features and perforinance of the Feedwater and Steam Systems are important elements In determining the ability of the operators to perform normal and abnormal operating evolutions in a .
manner which minimizes the number of reactor trips, challenges to safety systems and maintain the plant parameters within the post trip window. The actions associated with these systems are described in this section.
3.a Actions Associated With Emergency Feedwater Initiation and Control Rancho Seco developed a design and established a schedule for the implementation of an Emergency Feedwater Initiation and
. , m z z. . .i ,- . , Control-Systen:-(EEIch in responsesto:industryaleuons' learnFdf 5"* * " ' '
This system provides:
a) Safety Grade Auxiliary Feedwater initiation and control.
2 b) Safety Grade Atmospheric Dump Valve (ADV) control.
c) Safety Grade Main Steam Failure Logic.
The schedule for the implementation of this modification is currently Cycle 8. This system would have reduced the severity.
1 of the cooldown transient of December 26, 1985.
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4C-16 J
s The objective'of these actions are to assure EFIC can be implemented in a timely manner, and that interim operations prior to the installation of EFIC can be conducted in a safe' and reliable manner. Note, that the changes already described for control of ADV's, TBV's, and AFH control valves, provide .
control independent of the ICS, on loss of ICS power. .These are some of the functions to be taken over by EFIC. To achieve this objective, the following actions will be taken.
.1 Actions to be Completed Prior to Restart or Completion of j the Restart Test Program l
- 1. Implement control grade modifications to provide automatic steam generator level and manual flow control of Auxiliary Feedwater flow independent of
> ICS in the Contro1rRoom; m -- - -
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.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed
- 1. A detailed schedule will.be provided which describes the accelerated EFIC Installation, j.
i 4C-17
-.-- -,,-, , -.--- - - - .-y ,- ,.-- ,. ,y-,- ,--, - -- - - - , , , , - - - , - . - - .
3.b Action to Improve Auxiliary Steam Control Valve Operation on f\
4 Loss of ICS Power loss of ICS power during the December 26, 1985 event opened the auxiliary steam pressure control valve to mid-position, causing .
overpressure in the auxiliary steam header and lifting of one of the headers two relief valves.
The objective of these actions are to effectively address this plant design condition.
.1 Actions to be Completed Prior to Restart or Completion _of the Restart Test Program
^
Develop and implement plant modifications to the
-; 73-au x i 11 arys s te am con t rol szto c as surewri vescontrcrimm t-rm mne"n loss of ICS power.
3.c Actions to Reduce Main Feedwater Contributions to Reactor Trips i
The main feedwater system is a significant contributor to plant trips and transients. The objective of these actions is to reduce the number and severity of feedwater induced trips and transients.
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program 4C-18
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- 1. Review operation of Main Feedwater Pumps and status of incorporating lessons learned from October 2, 1985 event.
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- 2. Retune the ICS. .
- 3. Validate setpoints and proper initiation / interface with Auxiliary Feedwater System.
.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed
- 1. The recommendations contained in the.B&WOG Availability Committee Report, 47-1159449-00, "MFW
, ,, m ,, , ... _ - - - , .
, Pump-TripvReductiomProgram41nalxReporttW11lebe wWN : '*'"
evaluated for applicability-to Rancho Seco'and implemented as appropriate.
3.d Actions from Control Room to Ensure Capability to Isolate Main Steam System The two 50% capacity steam lines run from the Reactor Building to the Main Turbine. There are numerous penetrations into these lines which provide plant steam services and steamline protection. All service connections, greater than two inch diameter, are provided with_ remotely operated valves. The 4C-19
controls for these diesel power backed electrically operated
() valves are located in the Control Room. Safety and relief valves are not provided with upstream isolation valves.
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.1 Actions to be Completed Prior to Restart or Completion of ,
the Restart Test Program
- 1. Walkdown the Main Steam Lines and verify that each i service connection, greater than two inch diameter, is provided with capability to Isolate from the i Control Room.-
- 2. Tag control room switches to clearly indicate valves associated with A- or B-Steam Generators.
O._r...___. .. -
- 3. Evaluate MalnuSteam:Line: Supports'forceffects?# rex e'"**:-
flooding.
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- 4. Increase IDADS sample frequency for MSLFL Parameters.
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4C.4 EMERGENCY DIESEL GENERATOR RELIABILITY The reliability of emergency Olesel generators is important to the mitigation of certain events.
1 The objective of these actions is to enhance the reliability of the Rancho Seco Emergency Diesel Generator systems. To accomplish this objective, the following actions will be taken.
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program
- 1. Replace Turbochargers.
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.2 Actions to be Initiated as Promptly as Practicable, Schedule
-c,-- -- Deve l oped p Re sourc e s e Ass i gned and3Ma i nta i ned $ tin t i l:Compl e t~ed e** "M"
- 1. Evaluate performance history of the Bruce-GM Emergency I
Diesel Generators and develop recommendations for reliability improvement. l l 2. Determine and implement identified modifications.
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- 3. Enhance the preventive maintenance program.
O 4C-21
l 4C.5 REACTOR COOLANT SYSTEM AND PRESSURIZER O
d 5.a Upgrade Pressurizer Relief Valve Discharge Piping Supports 2 The subject piping was reanalyzed for dynamic loads .
Including 2-phase and 11guld flow as part of the response to TMI lessons learned. SMUD provided the results of this analysis and a justification for continued operation with the existing configuration by letter to the NRC on July 29, 1983. As committed, these supports are now being upgraded to restore their design margin.
.1 Actions'to be Completed Prior to Restart or ,
Completion of the Restart Test Program y-_ .. . _ . . . - . , -
- 1r Issue. ECNs forr.new.:-supportscandrsupportv"m"":"*" ' 'w modifications.
4
- 2. Inspect, reanalyze, and redesign (as required) ring structure anchoring supports to Pressurizer.
-3. Construct new supports and modify existing supports and ring structure (if required).
- 4. Construct pressurizer support structure modifications and modifications to existing work platforms reautred to resist new pipe support loads.
4C.6 ENHANCE THE POST ACCIDENT SAMPLING SYSTEM (PASS) OPERABILITY The objective of these actions is to enhance the operability of the PASS system to meet its design objectives.
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program
- 1. Complete the SCAS panel rebuild.
- 2. Complete associated peripheral equipment upgrades.
t 3. Document the compensating equipment in the enviro'nmental lab.
e-y - 4. Comp 1 e te work: . required sto 'solye tHj :moni toringeheat enmwrnmm ""-
tracing problems.
- 5. Revise operating procedures and complete training on revised system and conduct system functional test.
.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Comoleted i
- 1. Replace Dionex program controller.
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- 2. Install R-15044 Sample Dryers.
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.3 Actions-to be Programmed for the Longer Term i cO i
! Complete PASS decay heat valve replacement during the cycle 8 i
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4C.7 ACTIONS TO ENHANCE CONTROL ROOM /TSC AND NSEB HVAC - OPERABILITY AND N RELIABILITY Since the start-up and turnover of the essential HVAC systems for the Control Room /TSC and the NSEB during the 1985 refueling outage, .
several maintenance and operational problems with these systems have been identified. Most items had been identified and addressed by mid-December of 1985 by an interdepartment task force, with several additional items identified as a result of studies following the December 26 transient.
The objective of these actions is to ensure operability of these systems in accordance with the original design bases, improve reliability of the systems, and facilitate the maintenance and O operability of the systems.
.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program
- 1. Excessive HVAC noise affecting Control Rocm Operator communications
- a. Prepare anc implement a detailed action plan / testing program to identify source (s) and propose modifications.
O 4C-25
- b. Develop and implement those design and procedural
_, changes needed to reduce noise levels to allowable limits.
.2 Actions to be Initiated as Promptly as Practicable, Schedule .
Developed, Resources Assigned and Maintained Until Completed
- 1. Evaluate and. implement design changes if necessary to improve balancing capabilities.
- 2. Develop and implement the changes to install flow meters to facilitate surveillance testing of the Control Room /TSC HVAC filter units.
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- 3. Develop and implement the changes necessary to facilitate
. . . - . mai ntenance: ofrControls Room /TSC -HVAccequipmentM i MF;"*N"*""* '" --
replace Air Handler Unit access doors, modify the lube manifold to condenser fans, etc.)
- 4. Fire Damper upgrade actions
4
- b. Develop and implement the changes necessary to upgrade dampers la the wall between Contr61 Room and TSC.
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> 5. Control Room /TSC essential filter unit flow control i
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! a. Investigate flow control through filter units and l
recommend improvements.
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- 6. Essential HVAC compressor motors
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- a. Investigate replacement of existing motors.
- 7. NSEB essential air handler air flow
- a. Develop improved methods to adjust air flow.
- 8. If required, develop and implement design change to add a sample manifold to filter banks in each of two units.
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g 4C.8 INSTRUMENT AIR SYSTEM RELIABILITY Rancho Seco has experienced two transients in the past due to a loss, or partial loss, of instrument air. Shutdown to cold conditions is possible without the IAS but it is not a normal shutdown and may require the manual operation of important valves.
The objective of these actions is to improve the reliability of the instrument air system to minimize its contribution to plant transients and reduce the potential impacts to the cooldown transient.
.1 Actions to be Completed Prior to Restart or Completion of the' Restart Test Program d 1. Complete IAS system review to identify hardware modifications required toAmprove systersteliabilityirom wmm m :- ~
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- 2. Replace leaking letdown filter valve operators.
- 3. Add diesel-driven air compressor.
- 4. Provide bottled air backup to critical valves.
- 6. Develop and initiate modifications identified during system review.
I 4C-28
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. 2 Actions to be Initiated as Promptly as Practicable, Schedule i
' Developed, Resources Assigned and Maintained Until Completed ;
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- l. -Develop and implement modifications identified in IAS ,
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4C.9 REACTOR BUILDING PURGE FLOW RATE MEASUREMENTS The ability to determine containment purge flow rate is currently impacting the off-site dose calculations during containment purges.
The objective of these actions is to enhance this calculational capability.
.1 Actions to be Initiated as Promotly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Comoleted i
- 1. Perform engineering evaluation of flow measuring system deficiencies and identify appropriate modifications.
- 2. Install and test modifications.
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4C.10 : FIRE PROTECTION SYSTEMS
\
Rancho Seco has complete' d the Appendix R safe shutdown analyses and 4
has committed to specific additional modifications that are shown in the living schedule. -Additional changes and modifications which are .
desirable to improve the operability of the various systems have been identified as a result of the December, 1985 transient and the District's continuing efforts to improve plant performance and are
+
shown below.
4 10.a Fire Alarm Systems The objective of these actions is to enhance the performance of the fire alarm system.
.,;-- .1: Actions to ;be nInit: lated 'as eremptlyrapPractttabler5nermmnw'" "
Schedule Developed, Resources Assigned and Maintained' -
Until Completed
- 1. Develop and implement modifications to provide for manual operator override of a trip of the Auxiliary-Building ventilation fans.
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- 2. Upgrade the control logic and schematic diagrams for the fire protection system.
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! 4C-31 4
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- 3. Develop and implement appropriate upgrades to prevent I spurious signals on power transients.
- - l 10.b Separation of NSEB Damper Controls and Equipment The current control circuitry and equipment is to be upgraded to closecut an action item identified in the safe shutdown review and to support the operability of the NSEB emergency ventilation system.
.1 Actions to be Initiated as Promptly as Practicable, i Schedule Developed, Resources Assigned and Maintained until Completed i
1
% 1. Develop and implement fire alarm and HVAC panel 4- _ 1 c -upgradesatosepacateethe rcontroinctrewitryrandtrnnwr n n m -
equipment for the Train A and Train 8 Dampers.
10.c Water Leakage through Floors Following Actuation of Fire i Protection Systems The fire deteetion system in the NSEB Cable Shafts and Tunnels needs to be upgraded to close-out an item identified in the safe shutdown analysis and to support the EFIC system i
installation.
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l 4C-32
.1 Actions to be Initiated as Promptly as Practicable,
' Schedule Developed, Resources Assigned and Maintained Until Completed
- l. Identify vital areas of potential impact due to .
- leakage.
- 2. Evaluate effect of impact of potential leakage on safe shutdown equipment.
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- 3. Inspect all vital electrical equipment areas for-potential leakage paths.
- 4. Evaluate the results of the inspection and identify recommended corrective actions.
Develop and implement changes necessary to' address'
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5.
j- recommended corrective actions.
- 6. Review and upgrade as necessary preventative i maintenance procedures to maintain drain lines clear of obstructions.
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, 4C.ll MOTOR OPERATED VALVES A near term program is to be implemented to assure all applicable HPI and AFW motor operated valves maintain their environmental qualifications and have valve operator switches selected, set, .
- tested, and maintained properly. Key applicable features of this program are'to be extended to other motor operated valves in the longer term. These commitments will meet the. requirements of IE 4 Bulletin 85-03 and IE Notice 86-03.
i The objective of these actions is to assure reliable operation of i
motor operated valves.
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.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program
] s.
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- 1. Complete applicable portions of'the District commitments 1
to the NRC in accordance with Mr. John E. Ward's (SMUD)
I letter JEW 86-023 of 5/16/86 to Mr. J. B. Martin (NRC).
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.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed i 1. Complete the balance of the Olstrict commitments per
' letter JEW 86-023.
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l' i .3 Actions to be Programed for the Longer Term t6
- 1. Extend the motor operated valve program to include other l
j motor operated valves.
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4C.12 CRITICAL PUMPS FAILURE ON LOSS OF SUCTION During the December 26, 1985 event, an operator error resulted in both sources of suction to the make-up pump being closed resulting in major pump damage. .
The objective of these actions is to effectively address this issue to avoid reoccurrence.
.1 Actions to be Comoleted Prior to Restart or Comoletion of the Restart Test Program
- 1. Evaluate procedures and provide training to prevent recurrence.
k t 7, , n, -
.2 Ac ti on s to. be I n i t i a te d. as.Jromptl y a s, Prae ti rabl e ,r Schedu l e tt aws weeve ' (
Developed, Resources Assigned and Maintained Until Completed ~"'
- 1. Engineering is to review design philosophy for suction valve interlocks and alarms on critical pumps and identify appropriate modifications.
4C.13 MAINTENANCE PROGRAMS AND ACTIONS The quality of the maintenance programs has a direct impact on the material condition and reliability of-systems and components throughout the entire plant. The actions described below are .
Intended to provide 01 strict Management with assurance that the material condition of Rancho Seco's safety systems, and those systems required for normal control as well as post-trip control, are such that safe operation may be resumed.
.1 Actions to be Completed Prior to Restart or Completion of the 1
Restart Test Program l
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i 1. Inventory Calibrated Test Equipment (CTE) and calibrate and/or control use to prevent use of uncalibrated CTE.
- 2. Assure current calibration of all- in-plant ~lnstrumentation' used in the performance of surve1110nce testing.
- 3. Rework the makeup pump and return to service.
- 4. Complete the in-progress battery replacements (A, B, C, D, E, F).
- 5. Perform refueling interval surveillance of snubbers.
- 6. Complete rework of terminations in the Bailey Cabinets in j the Control Room (NNI/SFAS/RPS/ICS).
_- - ~ _ _ _ _ _ _ _ _ _ . . - - _~-__- _ -~ _ _ _ _ . - - _ _ _ , _ _ . _- . . - _
- 7. Perform biennial Diesel Generator Inspection and replace !
- 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 1 and steam generator heat removal applications (143), plant instrumentation required for survelliances, safety related HVAC and the Control Room normal HVAC system.
- 9. Complete Preventive Maintenance (PMs) on manual valves selected due to their functional position, e.g., Isolation-
%rm . . . , - .. , . of active > equipmentusucbas. pumps:scooteol.valvest heaten nanu wr exchangers, cross-ties.
.. 10. Repair valves investigated as a part of December 26, 1985 event troubleshooting. Includes FV-20527, FV-20528, FHS-063, FWS-064.
.2 Actions to be Initiated as Promptly as Practicable.
Schedule Developed. Resources Assigned and Maintained Until Completed
- 1. Develop a departmental procedure hierarchy and writer's guide for Maintenance Procedures.
- 2. Identify and prioritize procedures for generation and/or revision.
- 3. . Achieve authorized staffing levels within the maintenince organizations. .
- 4. Develop and/or revise the required programmatic procedures for the PM program to: assign
!i responsibilities, authority and accountabilities for the program; establish criteria and define the scope of the program; and define the interface with other work control processes.
- 5. Review existing PM tasks and frequency for critical i
[ equipment. Revise and augment as required by j programmatic, selection cr. iter.tausuuam nvuwe " -
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4C-39 U..._._.____.___,_,__.__,__.__.__,___._.___-____.-..*____,._.._._.,.._.
4C.14 ONCE THROUGH STEAM GENERATORS (OTSG'S) o Since Rancho Seco is a Pressurized Water Reactor (PWR), the staff
,i pays significant interest to the integrity of the OTSG tubes, whose condition can be inferred by analyzing secondary plant steam / water .
for traces of reactor coolant. Following the December 26 event, prolonged analysis was unable to conclusively establish whether or not a small leak existed within the OTSG's. As a result it was determined to take advantage of the cutage to identify any degraded l tubes and to take corrective action as required. Helium leak testing 4
i was done on both OTSG's which identified two tubes with very small leaks in the B-0TSG. Consequently, it was prudent to perform
! eddy-current inspection of the OTSG's to minimize the probability of 4 a tube leak following restart. The program was comprehensive with approximately 5,000 and 7,000 tubes being inspected in the A and B
..w m . 0TSG's, respectivelym0fetheseatubeen12dnathe M-0T5Geand252cinnthe ?nem+ r
- B-0TSG were plugged. Since the majority of'these-' tubes were-located'~ '
in the " lane" regions of the OTSG's where previous programs had plugged tubes, B&W (OTSG Manufacturer) recommended that the tubes in
, the " lane" region of both OTSG's be " sleeved" to minimize the probability of future leaks which would require down time and outaces for plugging.
.1 Actions to be Completed Prior to Restart or Completion of the
- Restart Test Program l
3
/
(
4C-40 1
k
- e i
h
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- 2. Perform Eddy-Current inspection of OTSG's to recommend ,
1
- tubes for plugging.
- 3. Plug those tubes identified in 1 and 2 above. j I
j 4. Develop program and licensing documents necessary to I sleeve tubes in lane region of OTSG's. -l i
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! 5. Install sleeves in selected tubes.
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I
.40 SYSTEM REVIEW AND TESTING PROGRAM i
1:
Rancho Seco has instituted a Systems Engineer program modeled after i an associated INPO Good Practice. Individual engineers are assigned.
specific systems for which they have the responsibility to know,
- among other things, the design basis, system limits and precautions, I
j and to monitor the system condition. In addition, they aid other, engineers,~ technicians, trainers and operators when interfacing with j.
l~ an assigned' system.
I j For the purposes'of the Action Plan, the-system engineer receives .
copies of all recommendations from the RRR8 which are relevant to
- . their assigned system. The Systems Engineer considers the recommendations in light of'their effect upon the system and presents-
{ an integrated system solution to the PAG. In particular, a review of j the systems functional basis, trends of previous maintenance history, ,
i
- . and test records will be conducted to assess the applicability of the recommendations to the issues associated with the systems ' performance l
I history. From this review it is expected that further l recommendations for system enhancements may be issued,. including recommendations for specific testing. Testing is expected whenever it is determined that prior testing may not have. adequately tested a j particular system design feature, or that a subsequent modification
! or functional requirement may need further testing or integration-demonstration.
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40-1 i
l t.-.-.._-_-......-.. - - . - _ _ _ . - - - - . . - . . . . - - . - . . . - , . - . - . , . . , . . -
E 4 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 qualified representatives to the Davis-Besse and Three Mlle Island sites to review their startup testing 1
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 test program are being reviewed for applicability to the Rancho Seco SRTP.
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4D.1 SYSTEM REVIEW AND TESTING PROGRAM 1.a Purpose .
The Action Plan incorporates a System Review and Testing 4
Program (SRTP) whose objective is to demonstrate that systems Important to safety are ready to perform their required
- function when Rancho Seco is ready to return to power. The.
questions this plan was developed to answer are:
t Which systems should be reviewed prior to restart?
What is the scope of the review?
What should be the criteria for the development of j appropriate tests to be performed?
What is the scope and detail of the testing program?
1.b Organization P
The SRTP Organization is shown in Figure 40-1 l
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! 40-3 I
l O O O RESTART l RESTART TESTING IMr cEMENTATICN MANAGER i
SYSTEMS REVIEW AND TEST PROGRAM DIRECTOR TEST REVIEW ASST.SYS. REVIEW GROUP AND TEST PROGRAM DIRECTOR l @ l l l l l l l
1 EACTOR FUWT STEN 1 PLANT POWER & AUXILIARY NODIFICATION SR. SURVEILLANCE SYSTENS SYSTEMS CONTROLS SYS. SYSTEtB TEST EACTOR & ISI SR. EtOIEER SR. EtGINEER SR. EtGIE ER SR. EtGIEER SLPERVISOR ENGIFEER StPERVISOR i
SYSTEtt SYSTEN SYSTEM SYSTEN START-UP REACTOR SURVEILLAFCE l EtEIT ERS EtGItEERS ENGIEERS ENGIE ERS EtEINEERS ENGIEERS SCEDULER
System Review & Test Program EiGIEER Organization l
1-1
h 1.c- Responsibilities
!g
} .1 System Review and Test Program Director 4
[ The System Review and Test Program Director is_ responsible i' to the RIM for the development, execution and summarization of the required program to assure component, j system and plant material conditions can support safe and
! reliable operation. In'this capacity he works with the Nuclear Department Managers and the Outage Manager, and 2
1 their designees, to ensure that the system reviews are thorough, test objectives are proper, and that the test
! procedures are developed, planned, performed and evaluated i
l In accordance with existing plant procedures. This position will work closely with'the Outage Manager to.
develop and maintain a detailed. schedule of work i
j activities in support of the test program.
i i
j The primary purpose of this position is to structure a l comprehensive system review and test program, including:
)
i 4
4 Defining the necessary organization and resources to accomplish the objectives.
1 i
Developing administrative controls.for the system-review and test program. As a minimum, these f administrative controls will address 40-5
responsibilities and authority, scope of system I
\ review, test procedure requirements such as review and approval cycle, content, format, etc., rules for 1-test conduct, and the evaluation and reporting of
- results. Where practicable, existing procedures will be used, e.g., AP.2 for the development and review of procedures.
I j
Defining the test objectives for those component's, systems and integrated systems to be tested.
i 3
Defining the test methodology to be used to accomplish the test objectives. This includes a l
compilation of the procedures to be used, and which
- procedure is being used to satisfy each test s
objective.
Preparation of test procedures.
J Defining training requirements for test personnel.
t Preparation of test results packages.
i Preparation of schedules for procedure development and test implementation in concert with the Outage j Manager.
O j 40-6
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. The Test Program Director is expected to provide the RIM i
with weekly status reports.
.2 Test Review Group The Test Review Group shall be established as a Task Review under Section 3.k of the Plant Review Committee l Charter.
I
'The Test Review Group shall contain the following members:
1 Nuclear Plant Department, Chairman 1
Nuclear Engineering Department Member Babcock & Wilcox Representative Member l
Bechtel Corporation Representative Member Quality Assurance Department Member i
T The System Review and Test Program Director will nominate j members for the Test Review Group. The ' Plant Review Committee 1-
! will confirm the members. The responsibilities of this Test i
j Review Group include the following:
4 Review of the Test History Review section of the. System .
l
! Status Report to confirm that the previous and proposed I i
testing will demonstrate all system functional
! requirements.
1 40-7 i ,
i
i Review of all Test Specifications to assure that the scope
~
4
'\ and methodology demonstrate the. system functions.
Review of related Special Test-Procedures and new or revised Surveillance Procedures and recommend approval to the Plant Review Ccmmittee.
Review of all restart related test results. ;
i i
1.d System Selection for Systems Review In order to expedite the' process of systems review, it was decided 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 (PP&MIP) process.
The systems selected by Toledo Edison for'their Davis-Besse j restart program were chosen as a base for 'the initiation of the systems review process. It was recognized that a different list of systems warranting a review would be identified as the PP&MIP process is concluded. Additionally, the PP&MIP process l could conclude that some of the systems chosen by Toledo Edison were not appropriate since these Toledo Edison systems do not ,
relate to problem areas identified as applicable to Rancho Seco. '
4 1
40-8
- - . - . . --. ,-~, .- , . , . . . - - . . - . - - - - - . - . - , - - - - - . - . _ , . . .-
4 Based on the above considerations the following criteria form i
the basis for the selection of systems for review:
I
- l. Has had a known history *of significant or recurring problems (
5
- 2. Was related to the 12/26/85 Event
- 3. Is being significantly modified i
2
- 4. Has significant potential for initiating or adversely affecting plant transients. I l
3 The Systems Review and Test Program Director shall propose and the PAG shall approve the list of selected systems.. A tentative listing of such systems is as follows:
i i
- 1. Reactor Plant Reactor Coolant System Decay Heat Removal System High Pressure Injection System Seal Injection and Makeup System Purification and Letdown Systt-m Nuclear Service Cooling Water System j Nuclear Service Raw Water System ;
i 40-9 i
. . . . .. . . _. . --. ._ . .. - - -. . _ = . . . . . . - _ - - - .- - - . . . -
- 2. Steam Plant A Steam Generators System '
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- Main Steam System h Main Feedwater System '
I t
. Auxiliary Feedwater System a
l j 3. Power / Control Reactor Protection System (including ARTS) !
l Safety Features Actuation System i
- Main Steam Line Failure Logic System i
j Integrated Control System i i
Non-Nuclear Instrumentation System
- l 125 Volt DC Vital Power System 125 Volt DC Non-Vital Power System 120 Volt AC Vital Power System i 480 Volt AC System 4160 Volt'AC System l 1 6900 Volt AC System '
i l
- 4. Aux 111ary.
Control Room /TSC Essential HVAC System 4
Emergency Diesel Generator System l
. Component Cooling Water i
Plant Air System Instrument Air System l l 40-10 i
I Based on the quantity and significance of issues raised by the PP&MIP process, other systems may be designated by the PAG for systems review.
It is emphasized that the ultimate 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 I
history as well as a review of industry experience and other relevant sources. The system related output of investigative groups of the PP&MIP are fed through the !
RRRB to the appropriate System Engineers. On the basis of ,
4 this evidence the System Engineer may Identify additional 1
systems requiring a review.
j The decision to perform additional system reviews or to terminate a system review will be approved by the PAG.
! 1.e System Review Summary r
.1 A systems review will consist of the following:
- O 40-11 l
i _ __ _ _ ._ _ -
r-p 1. The system functional description will be compiled <
l using the appropriate technical documents.
l 2. Original ' plant startup and subsequent Surveillance j and Special Test records of these selected systems 1 :
} will be reviewed to determine whether all of the 1
! functional characteristics have been satisfactorily demonstrated by test.
1 j 3. All related PAG recommendations and corresponding
- j. dispositions will be assembled and reviewed.
j :
l 4 i I l 4. The modifications and test history for the system will be reviewed to determine the impact of ;
modifications on system functions and the adequacy of previous testing.
- 5. Confirmation that IE Bulletin 79-14 walk down was performed for safety related systems.
I
- 6. A material condition walk-down will be performed.
I:
- 7. Results of the Davis-Besse test program will be
- reviewed for applicability.
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l
.2- System Functional Description
- 1. Definition
- a. System Functional description is a listing of the capabilities that the system must provide in order to assure reliable operation and effective accident mitigation.
1
- b. Generation of System Functional Description The System Engineer will prepare a System Functional Description. Source documents to be used are NEP 5400 (System Design Bases), the USAR, Technical Specifications, B&W Guide Specifications, Bechtel Rancho Seco Design Manual, vendor manuals, and applicable design
- calculations as appropriate.
The System Functional Description will be l documented in the format defined in Figure 40-2.
.3 Test History Review-4 The system engineer will review the testing that has been performed on the system. :A comparison will be made to ,
determine if all of the functions defined in the System l Functional Description have been demonstrated by testing.
40-13 1
.. - - . .- - . - . - . . - - 1
N Test History includes original Startup Tests, Surveillance i Tests, Post-Modification Tests, Post-Maintenance Tests, and Special Tests.
The modification history of each system will be reviewed
] to determine if modifications have occurred after any of the. functional tests that would invalidate previous functional tests.
i If'the functional testing is found to be inadequate, additional testing will be specified.
The results of the test and modification history review will be documented in the System Status Report. Copies of tests used to demonstrate system function will be included or referrenced.
.4 Review of Recommendations and Resolutions-The RRRB will forward copies of all systems related recommendations with the resolutions to the System l
Engineer. The System Engineer will consider the combined effect of the solutions being applied to the system and determine if all known system deficiencies are corrected.
He will also satisfy himself that no new issues are created.
4 The review of recommendations and dispositions will be documented in the System Status Report.
40-14
, .5 Review of Test Results from Davis Besse The Systen 2ngineer will review the system / component problems identified at Davis Besse through the Davis Besse test program for applicability at Rancho Seco. l The Davis Besse test results review will be documented in
, the System Status Report.
- 6. Review of Recent Maintenance Activity and Maintenance History Trends.
i The System Engineer will review trends of maintenance history for indications of system deficiencies.
The System Engineer will review system maintenance that has occurred since the 12/26/85 transient for adequacy of 4 post-maintenance testing. If post-maintenance testing is
- found to be inadequate, additional testing will be 4
specified.
The System Engineer will document the maintenance activity and the maintenance history-trend reviews in the System Status Report.
)
40-15
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{'s .7 Material Condition Walkdown l The Systems Engineer together with a designee by the Operations Manager will. Inspect the installed system for evidence of a deterioration in material conaltion. A check list will be provided in QCl-12 to assure consistency in the conduct of these reviews.
.8 Operability Finding Based on the testing history review, the review of recommendations and dispositions, and the recent-maintenance activity, the Engineer will specify any additional maintenance, modifications, procedure changes, and testing that.ls needed to assure reliable system operation.
He will state that upon completion of the defined work scope and his additional requirements (including testing) the system is ready for operation.
.9 Independent and Management Reviews The Test Review Group will review the Test History Review section of the System Status ~ Report. They will satisfy themselves by review of submitted test procedures and
[~ results, that the system functions have been demonstrated.
40-16 1
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The PAG will review and accept the System Status Report. ,
l.f Testing Program i
i
.1 Based upon the Systems Review, the Systems Engineer will :
, recommend a testing program that addresses:
- 1. Testing Due to Modifications
- 2. Post-Maintenance Testing
- 3. Vc>rification of Safety System Operability f
! 4. System Function t
f
- 5. Special Testing Where the PP&MIP and System Review 4
Indicate: ,
i
., a. Original functional requirements are not assured because of subsequent modifications or identified problem areas.
I
- b. Reoccurring system-related problems.
Note: The District has already determined that Loss of ICS, NNI and Instrument Air Tests will be conducted.
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- ec - ev- - se e- g- m e - -tmy 7- - >---g ~-e
c 6. Routine testing done during startup O
.2 It is planned to perform the following periodic tests.
- 1. Integrated Leak Rate Test of Reactor Building
- 2. Complete the Balance of Ten-year In-service Inspection (except for those inspections requiring removal of the reactor vessel head).
- 3. Integrated Engineered Safeguards Actuation Test j
- 4. Emergency Olesel Generator Bienn11 Inspections
.3 To confirm the adequacy and effectiveness of the Action Plan process in identifying and resolving system deficiencies. A sample system which would not require testing under this process will be selected by the PAG.
This system will undergo an appropriate system functional test.
Tables 1-A through 1-C describe the tests-identified to date. Normal testing associated with the plant performance monitoring program has not been included.
These tables will be modified as additional tests are identified in our ongoing review program.
40-18
For each test, a test specification and detailed procedure O will be developed. Examples of the test specifications j
l are shown in the Appendix. j l
i In order to put the extent of the test program into perspective, the District has performed it's own assessment of the Davis-Besse, Three Mile Island Unit 1, and the Rancho Seco test programs.
, The System Review and Test Program Director will develop a training program for test engineers. This prograd will cover i
the generation of Test Specifications, the writing of procedures, the revision of procedures (both permanent and temporary changes). test conduct and the review of results.
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4 TABLE 1-A PAGE 1 RANCHO SECO RESTART TESTING MAJOR TEST OBJECTIVES: Verification that the system and/or component that was. subjected to a design change performs in accordance with the modified design and is properly integrated into overall plant operations. .
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD L
Nuclear Fuel Core performance Verify core.is operating within design assumptions for reactivity Startup physics and data. test procedures
- STP-252, STP-253, 4
SP 209.01 '
Verify boron rundown curve is applicable to cycle 7. Adjust predicted curve, if appropriate, using' steady state HFP data.
Verify proper operation of the incore detector system. Re-run portions of BOC-7 startup l physics testing.
Letdown and Letdown filter Test valves installed or worked on to ensure operability and to Special Test Purification pneumatic valve determine operating characteristics (to include stroke and time, Procedure leak test, and function test).
Auxiliary ICS~ Control To verify that the design changes made to the AFH valves control Special Test feedwater Valve - scheme for loss of ICS DC power were installed properly and the Procedure operability design intent was achieved. .
Reactor System function To verify proper calibration and correct trip capability. To SP 200.08A, B, C, D Protection' calibrate and test all. input devices to the RPS. (None of the SP 200.128 System- changes in this outage change the function of the RPS. Therefore,
.: the regular refueling outage system tests will provide complete'
- i. function testing of the RPS.)
Safety System function To verify proper calibration and test of all channels of SFAS to SP 200.13, 203.01A, features test actuation of all SFAS channels (none of the. changes in this B 203.04A, B l Actuation outage change the function of the SFAS. Therefore,-the regular refueling outage system tests will provide complete function testing of the SFAS).
mo m .
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O O O TABLE l-A PAGE 2 RANCHO SECO RESTART TESTING MAJOR TEST OBJECTIVES: Verification that the system and/or component that was subjected to a design change performs in accordance with the modified design and is properly integrated into overall plant operations.
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Non-Nuclear System function To verify that the termination and rework program did not result in Maintenance Instrumenta- Improper terminations by verifying proper relationships between Instruction tion INPUTS, PROCESS PARAMETERS, AND OUTPUTS.
Integrated Valve function To verify that the design modifications made to the ICS and its end Special Test Control System with loss of devices (TBVs, ADVs, and AFH control valves) were installed properly Procedure ICS power and the design intent was achieved.
System To verify that the termination inspection and rework did not result Special Test Functional Test in improper terminations by verifying proper relationships between Procedure INPUTS, PROCESS PARAMETERS and OUTPUTS.
Annunciator To verify that the design change on the interface between the ICS Special Test verification ~ sources and the Annunciator System were installed properly and the Procedure design. intent was achieved.
Power supply To verify proper operation of power supplies. Maintenance function Instruction Indications on To verify control board indications on loss of ICS DC power. Special Test loss of ICS Procedure Module To ensure accuracy of process instrumentation. Normal calibration calibration Tuning at 10%, To verify that the ICS control loops perform in accordance with the Special Test 40%, 75%, and modified design and is properly integrated into the overall plant , Procedure 90% power operations.
O O O TABLE l-A PAGE 3 RANCHO SECO RESTART TESTING MAJOR TEST OBJECTIVES: Verification that the system and/or component that was subjected to a design change performs in
, accordance with the modified design and is properly integrated into overall plant operations.
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD DC Power Battery capacity Verify the new plant batteries are capable of supplying their design Special Test loads. Procedure Control Room System function Verify proper function of overall system. Special Test Essential Noise level Evaluate effectiveness of noise reduction mods; Procedure HVAC Cooling capacity Verify capability of system to provide adequate cooling during all Special Test expected ambient temperature conditions (cold and hot weather Procedure conditions).
System integrity Verify system integrity by performing leak testing. SP 211.018, 211.01E Instrument. Diesel driven Auto start' test and time. Check for clean dry and oil free air. New Special Test Air air compressor Monitor motor and compressor parameters. Pressure and flow rate of Procedures acceptance supplied air. 8-hour endurance test. Full break System integrity in as per manufacturer. Post test desicant and filter change out.
Engine and compressor service after test as required. Monitor charger and block heater.
Back-up gas N2 bottles, bottles function test, timed test for four hours, stroke count test, time -
operator to determine maximum startup time if required.
Fire Auxiliary To verify that detection and suppression systems (zone 110) were STP-949 Protection feedwater pump installed properly and that the design intent was achieved, and to Hater deluge system establish baseline operability in accordance with T.S. 3.13.1 and acceptance 3.14.3.
To verify that piping modifications to suppression syst.em piping in Special Test zone 104 were installed properly and that the design intent was Procedure achieved. '
1 1 r J ,
- TABLE l-A i PAGE 4 RANCHO SECO RESTART TESTING MAJOR TEST OBJECTIVES
- Verification that the system and/or component that was subjected.to a design change performs in accordance with the modified design and is properly integrated into overall plant operations.
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD PASS System function To verify that the design modifications made to the PASS were Special Test .t installed properly and the design intent was achieved. Procedures l
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TABLE l-8 l PAGE 1 RANCHO SECO RESTART TESTING NAJOR TEST OBJECTIVES: Verification of operability of safety systems and functionality of non-safety systems following.
maintenance activities.
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Reactor System integrity Verify by Operational Leak Check and system walk-down, that the RCS SP 207.02A/B Coolant System Pressure Boundary is intact as required by Technical Specifications. SP 200.03, 207.048 Verify setpoint of Pressurizer Code Safety Valve SP 207.03 Verify RCS High Point Vents functional by Flow Test- New Surveillance Procedure System Verify instrumentation function. SP 200.14, 209.09 Instrumentation
- l. High Pressure Make-up pump Verify proper operating characteristics of the Make-up pump Special Test Injection performance following rework. (pump curve) Procedure.
- Make-up Containment' Penetration leak . Verify proper closure and pressure tightness of individual SP 205.02 Building rate penetrations.
Steam Integrity Verify by system walkdown that OTSG pressure boundary is intact. STP Generators Main Steam Relief valve Verify relief valve setpoint correct for PSV-20549. WR 111478, MT 006 i setpoint ~
Verify acceptable operation of acoustic monitors. Perform I.036 l
HR 111477 Reactor *
' System Function To verify proper calibration and correct trip capability. Surveillance Protection To calibrate and test all input devices to the RPS. Procedures 200.08A/B/C/D 200.128 Safety System function To verify proper calibration and test of all channels of SFAS. Surveillance Features -To test actuation of all SFAS devices on all SFAS channels. Procedures 200.13 203.01A/B.
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4
. _ _ . - _ _ _ _ _ . - _ -__.__..___.____._____m_ _ . _ _ _ _ _ _ . _ _ . _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ m w - .-
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TABLE l-B PAGE 2 RANCHO SECO RESTART TESTING MAJOR TEST OBJECTIVES: Verification of operability of safety systems and functionality of non-safety systems following maintenance activities.
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Integrated Continuity checks To verify functionality following the term 1 nation, rework and Maintenance Control Functional preventative maintenance activities by performing continuity checks Instructions check PSM Si and for all new wire wraps.
S2 trip times To verify functionality following replacement of S1, S2 and the Halntenance Power Supply Monitor by performing a trip time measurement. Instructions Non-Nuclear System function To verify functionality of the NNI system following the termination STP Instrumenta- rework and preventative maintenance activities by system operational tion checks.
Control Room System Function Verify function of component and overall system. Special Test Essential HVAC Procedures SP 211.01A/D Auxiliary Charcoal Verify integrity of carbon and HEPA filter banks. .SP 211.04B Building HVAC performance -
Auxiliary Bldg.
Exhaust Fans
. _ _ . _ _ _ . ~ _ . . . . _ . . _ __ . _ . _ _ _ _ _ - _ _ _ _ . _ _ . _ _. .
~
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- TABLE l-C PAGE 1
, RANCHO SECO RESTART TESTING
, MAJOR TEST OBJECTIVES: Verification of system operability per technical specification.
I- SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE /METH00 i
i Core Flood . Level indication To satisfy ISI safety evaluation of 9/25/84 on full stroke testing SP 200.02 A.4 4
check valve of CFT check valves CFS-001 and CFS-002 SP 200.14 AP.103
- 2. New Surveillance.
Procedure to be
, written 1
! HPI HPI pump ' Test operability of HPI and M/U pumps. Test and inspect system SP 203.02A/B/C performance; HPI valves. Pump venting. SP 203.03 valve performance SP 203.02H/I/J Decay Heat Pumps; valves Pump performance test. Vent pumps. SP 203.05F/G system integrity SP-203.05A/B
' Stroke an'd time motor operated valves. New Surveillance Procedure 203.06A/B
' System leakage test. SP 205.02 l l SP 203.09 '
i Ultrasonic
! Containment Pumps Pump venting. _
SP 204.01A/B,
. Spray Valves Stroke and time motor operated valves 204.03A/B' j Sodium hydroxide System. leakage test SP 204.01C/D/E,
, concentration 204.07 '
System integrity New Surveillance procedure AP 306 VI i e A.7 AP 105 A.7 6
6 p 4
p % m .
U TABLE 1-C PAGE 2 RANCHO SECO RESTART TESTING 4 MAJOR TEST OBJECTIVES: Verification of system operability per technical specification.
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Nuclear Pump performance Pump performance test SP 203.07A/B Service Cooling Hater Nuclear Pump performance Pump performance tests SP 203.07C/D Service Raw Hater' Borated Hater CBAST concentra- Verify CBAST boric acid concentration SP 200.10A tion Pump performance test SP 210.llA Pump performance I-022 Valve Performance Verify system check valve performance SP 210.'11B Containment Integrity LLRT. underway; ILRT scheduled prior to startup. SP 205.01 Building (penetrations) Perform containment SP 205.02 SP 205.04 Spent fuel Pump performance Verify pump flow, bearing temperature and differential pressure. SP 210.10 Cooling
~
i Leakage Test Steam Integrity Verify primary to secondary leakage less than 1 gpm. Mechanical Generators Engineering j Guidelines i
ISI examine 12,000 tubes using eddy current examination Ultrasonic Eddy Current Main Steam Valve performance Verify 17 of 18 MSSV's operable by walkdown. Plant Operations Verify 1 TBV or 1 ADV operable per OTSG by walkdown. Procedure B.2,
" Plant Heating and Startup" 3.11.1.2 and 3.12.1
O V TABLE 1-C
'PAGE 3 RANCHO SECO RESTART TESTING HAJOR TEST OBJECTIVES: Verification of system operability per Technical Specifications.
SYSTEM SCOPE' SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Main Steam Verify Main Steam Safety Valve Setpoints. SP 210.02 (cont.) Verify performance of Main Steam Isolation Valves SP 210.02A System Integrity Hydrostatically test one loop. Augmented ISI Auxillary Pump performance To verify that each Aux 111ary Feedwater pump starts as designed SP 210.01A/8, feedwater automatically upon receipt of each Auxiliary feedwater actuation 210.02H test signal. Verify Pump performance per ASME,Section XI Valve performance To verify that each automatic valve in the flow path actuates to its SP 210.02C, 210.02H correct position upon receipt of each Auxiliary feedwater actuation SP 210.02I, 210.02J test signal and performance of check valves.
Flow Path To verify that the motor-driven AFW pumps can pump water from the. SP 210.01F CST to the steam generator.
Flow instrumenta- To ensure accuracy of flow instrumentation. Instrument tion Calibration Ultrasonic-Reactor Input device To verify proper calibration and correct trip capability. Su'rveillance Protection calibration To calibrate and test all input devices to the RPS. Procedures System 200.08A/B/C/D calibration 200.12B, 200.12C Safety System To verify proper calibration and test of all channels of SFAS. Surveillance Features: calibration To test actuation of all SFAS devices on all SFAS channels. Procedures Actuation Integrated test . 200.13 203.01A/B 203.04A/B 200.09
TABLE l-C PAGE 4 RANCHO SECO RESTART TESTING MAJOR' TEST OBJECTIVES: . Verification of system operability per Technical. Specifications.
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Nuclear- Calibration Verify calibration of Out-of-core monitors. .SP !!00.12A, I-103
- .Instrumenta- Vericy calibration of back-up recorders SP 100.04. 200.05 -
tion Non-Nuclear Calibration EH0V Power Position Indicator. SP 200.20 '
+
.Instrumenta- EMOV Block Valve Indication SP 207.03B
'tton Radiation Calibration Verify radiation monitor channel operability to include monitor SP 200.07 j Monitoring calibration and associated equipment operation, alarm or trip as SP 200.01 appropriate.
Verify RB and Auxillary Building Flow Rate Instrumentation .SP 452 PASS System function Verify the capability to obtain and analyze reactor coolant and Special Test Reactor-Building atmosphere samples under accident conditions. Procedures i Verify Instrumentation associated with PASS to be functional. -SP 200.01, 200.14
- , SP 200.108, 200.07 New Surveillance i Procedures Hydrogen System function Verify operability of containment hydrogen monitoring system, SP 201 08 Special-Monitoring including hydrogen analyzers Test Procedures Calibration, New Surveillance Procedures l
. Seismic Calibration To ensure the accuracy of the instrumentation. SP 200.06L Monitoring i
-Emergency Diesel Emergency Diesel Generators and Auxiliary systems functional test. SP 206.03A/B
' Diesel performance Perform biennial diesel engine inspection. SP 206.01A, 4 Generator 206.02A/B SP 206.01B l.
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! TABLE 1-C -
PAGE 5 RANCHO SECO RESTART TESTING i
MAJOR TEST OBJECTIVES: Verification of system operability per technical specifications.
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Control Room Charcoal Verify methyl todide removal efficiency of. carbon, integrity of Special Test Essential condition; filter carbon and HEPA filter banks, system integrity,-system function. ' Procedure HVAC condition; system SP 211.01A/D Integrity; system SP 211.01B/E function SP 211.01C Normal HVAC Filter testing Verify operability and function. STP and Plant-on selected units performance Functional test monitoring program
- on selected units RB HVAC System function Verify system starts and operates properly. SP 211.05 RB Normal Cooling Units System Function Verify system starts and operates properly. SP 204.04A/B of RB Emergency Verify operability of RB Emergency Coolers. . SP 204.02 Cooling System Verify integrity of carbon and HEPA filter banks and system SP 211.03A/B efficiency.
System Function Verify methyl lodide removal efficiency of carbon, integrity of SP 211.02A/B/C of RB Purge carbon and HEPA filter bank, system integrity, system functioa.
Verify Reactor Building Purge Valves. SP 205.07D Fire Pump performance To ensure surveillance testing on all Fire Protection systems is SP 201.038/D/G Protection- System integrity current and satisfactory. SP 201.031/H Hater Detection system function .
- Carbon Dioxide Control' system Semi-annual surveillance should be performed prior to restart. SP 201.03D l function d
O O O TABLE 1-C PAGE 6 RANCHO SECO RESTART TESTING MAJOR TEST OBJECTIVES: Verification of system operability per technical specifications.
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Pipe Supports Snubber Verify hydraulic safety snubber operable by inspection and functional SP 201.10A/B inspections testing.
Emergency System Verify emergency 125V DC system functional including periodic checks SP 206.04 Power functional and discharge load testing. ,
Verify pressurizer heater. power from 2A2 and 282 buses. SP 206.06A/B i
1
Rev. 8
\ APPENDIX A DISTRICT BOARD OF DIRECTORS' POLICY STATEMENT ON ,
PERFORMANCE IMPROVEMENT AT RANCHO SECO I. INTRODUCTION On July 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 has 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 experienced one of the poorest overall performance records.
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, A-1
reliably and economically. Carrying out this mission in a responsible !
, manner involves continuously balancing the emphasis on safety, which is primary, reliability, and economics to try to maintain an optimum relationship. It includes strict adherence to the SMUD Quality Pro-gram. It also includas taking the necessary short-term actions to .
achieve and maintain a desired balance over the long-term. In carry-ing 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 meet-ing each of these responsibilities in an appropriate manner is signi-ficant. 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.
4 II. POLICY STATEMENT The District's Board of Directors is committed to achieving a prompt j and significant improvement in performance at Rancho Seco and to s provide the support necessary to achieve standards of excellence in A-2 l
i all aspects of nuclear activities. To reinforce that commitment, the Board has established the following objectives related to performance improvement at Rancho Seco.
h.
Near-Term District Objectives- . ,
Accomplish those actions which will substantially reduce the likelihood of another significant transient at Rancho Seco.
4 Initiate longer term actions which will contribute to 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.
I t Performance Improvement Program Goals i
i The 1990 goals which have been committed to INPO include the following
- items
i Improved Equipment / Plant Availability
- l Reduced Forced Outage Rate l
I l Reduced Reactor Scram Rate ,
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! Reduced unplanned Safety System Actuations l 3
Reduced Safety System Unavailability Improved. Thermal Performance i
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Reduced Low Level Waste Volume Reduced Personnel Exposures Reduced Industrial Accident Rate l
d III. POLICY IMPLEMENTATION AND PLANNING GUIDANCE . !
l The Board considers the Performance Improvement Program at Rancho Seco l 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 Fro-gram Plan which encompasses all projects related to the Performance Improvement Program.
The Board of Directors encourages District participation in the activ-itles 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 inter-actions with other utilities with plants of similar design.
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APPENDIX B 8
l SPECIFIC DISTRICT RESPONSES TO l NUREG-1195 FINDINGS f
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. APPENDIX B SPECIFIC DISTRICT RESPONSES TO ,
NUREG-Il95 FINDINGS The following are the District's responses to the findings and conclusions section of NUREG-1195.
FINDING - SECTION 8.1
- 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 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 of the failure of the power supply monitor by having an independent laboratory conduct additional analyses).
DISTRICT RESPONSE Specific plant modifications were engineered and installed to address the identified lessons learned. Engineering Change Notices and subsequent field work accomplished the following:
B-1
1eads to the power supply monitor now go directly to the power supply bus on the ICS and NNI.
Inspection and correction of terminations (i.e., lugs) was completed.
ICS Power Supply Cabinet Bus wiring was replaced.
the original power supply monitor, and the Sl/S2 switches, have been sent to an independent laboratory for analysis.
A new power supply monitor has been calibrated and installed.
The District voluntarily undertook a program to test and inspect the terminations throughout the ICS, NNI, SFAS, and RPS cabinets, supplied by Bailey Meter Company.
Engineering is investigating ways to improve the reliability of the power supplies for the ICS/NNI. Consideration is being given to redundant and/or independent power supply monitors. Pending completion of these investigations, no decisions have been made as to the nature or schedule implementation of proposed enhancements.
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t FINDING - B.2
- 2. Upon loss of ICS de power and the subsequent automatic repositioning of a number of valves in the plant, the design of the ICS also caused the loss of remote control of the 7- affected valves from the control room which necessitates manual actions locally at the valves.
DISTRICT RESPONSE 4
i I
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 automatically controlled to maintain correct steam generator l level on loss of ICS Power. The operator will be able to take I
manual control from within the Control Room.
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FINDING - B.3 C
- 3. An AFW manual isolation valve could not be shut by the i
operators 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.
i
.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 i
condition. As an element of the troubleshooting effort, the i
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 4
'B-4
i fs 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 i identified 143 valves which will be verified operable prior to resumption of power operation. These manual isolation valves are characterized by their purpose which is to allow isolation of important active equipment such as pumps, valves, and heat exchangers. They were selected to include both primary and secondary plant systems necessary to power production or nuclear safety. Functicn, not class, was the selection criteria. The program involved actual stroking of the valves and, where
, a necessary, servicing with lubricants, packing, or adjustments was done and documented. Statistics are being collected and evaluated to establish a summary status of valves in similar service.
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 detail.ing-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.
B-5
FINDING - SECTION B.4 4
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- 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.
OISTRICT RESPONSE O
s) During the December 26 event those actions necessary to respond to the consequences of the Loss of ICS pcwer were appropriately
, defined within the E0P'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 i
of ICS Casualty Procedure would have been helpful in expediting the power restoration, that function correctly should not be a part of the E0P's. The E0P's must address the condition when ICS Power cannot be restored, for whatever reason, and a i I
/"N specific Casualty Procedure has been implemented for this
.U requirement. i i
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A review of the ATOG determined that there were no other needs or requirements which were not incorporated in the E0P's.
i B-7
4 FINDING - 8.5
- Os
- 5. The EOPs at Rancho Seco direct the operators to trip the -
appropriate feed pumps to terminate flow if the feedwater
- flow cannot be isolated. This was not done during the
. December 26, 1985 incident. The operators were reluctant to i
stop the AFW pumps even.when they had difficulty stopping flow to the once-through steam generators (OTSG) by valve .i 4
operation. The 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 i AFW system by NRC and others, and a lack of confidence in the reliability of the AFH pumps (i.e., fear that the pumps would not resta'rt if stopped).
DISTRICT RESPONSE l1 j -
The E0Ps did not contain specific parametric criteria such as RCS Temperature, Steam Generator Level, or. Pressurizer Level for when to trip main and auxiliary feedwater pumps during an
. overcooling. Lack of specific criteria'let the operators be influenced by perceptions of NRC concerns-regarding aux.111ary feedwater operability. Procedures have been revised to specify p when to trip main,. condensate and aux 111ary feedwater pumps.
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8-8
1 r~'s 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. A detailed review of the starting and operatir.; reliability of the Auxiliary Feedwater pumps was done which did not support a lack of confidence in this equipment by the operators. In the twelve year operating history (315 attempted starts) three instances of failure to start these pumps were noted.
4 I
In the first case, in early 1975, 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 third O'-' occurrence was a failure to start P-318 on its turbine drive, also for a surveillance test. P-319 has always started. There has.never been a time when there has-been loss of function of the Auxiliary Feedwater Pumps.
B-9
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FINDING - 8.6
- - 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-accordance with their E0Ps). Their training and procedures were
! not adequate to resolve this conflict and to some extent tended to provide conflicting indications of the appropriate priorities.
DISTRICT RESPONSE
- The Emergency Operating Procedure (EOP) in place on December 26 included rules which specifically state the events or plant D conditions which mandate throttling of HPI flows. Under Section 2.2 it is clear that HPI flow should be throttled to prevent exceeding i
brittle fracture limitations, i
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
f
- 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.
O-i B-10
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- 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.
2 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 complex events.
Although improvements in the ATOG derived 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 improve:nents have been directed toward precluding conditions which precipitate PTS issues.
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FINDING - B.7 l
- 7. The operators received neither classroom nor simulator training on the overall plant response to either the total loss of ICS de power or the restoration of ICS de power.
DISTRICT RESPONSE Training programs have been revised, to address the lessons learned from the December-26 event, and expanded to include the plan't 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:
Emergency Operating Procedures (EOP) Training including all I
steps necessary to terminate overcooling, or OTSG overfill from i
any cause, including the loss of ICS power.
Effect of changes to ADV, TBV, and AFW valve operation following loss of ICS power.
Command and control training, including implementation of the Emergency Plan.
B-12
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f'-'s Recovery from SFAS, i.e., restoring normal makeup and letdown
! flow. l Differences between the B&W Simulator and the facility (Operator 4
traps).
HPI and AFW throttilng and pump trip criteria.
PTS recovery actions.
Cooldown rate interpretation and tracking.
Conversion from AFW to MFW flow.
The District 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 effects of Restoration of ICS Power are under investigation by i the B&W Owners Group. Modifications to ADV's, TBV's, and AFW Control l Valves mean that following loss of ICS power, control will
- automatically transfer to independent controls. Procedures cause 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. ~0nce stable ICS conditions are observed .
4 the operator can return each device to ICS control.
B-13 y
,~ FINDING - B.8 i
- 8. The operators who investigated the loss of ICS power did not )
adequately understand the ICS power system configuration. When 120 Vac power is still available from the IC bus and the ICS dc power supplies de-energized, the most credible cause for the loss of ICS de power was the opening of switches Si and S2.
However, the operators did not recognize this fact and, as a result, did not shut the switches until 26 minutes into the transient. The fact that several operators did not recognize 1
that sw-itches S1 and S2 were OFF suggests that their training on this crucial system was not adequate. In addition, although simplified drawings of-the non-nuclear instrumentation (NNI) s power supplies were posted on the NNI cabinets, comparable drawings for the ICS power supply had not been provided.
DISTRICT RESPONSES a
Since the event Training has been conducted on the design and operation of the ICS/NNI power supplies. In addition, S1 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. 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 staote plar.t condition, following Loss of ICS Power.
D B-14
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- FINDING - B.9 l
c 9. It does not appear that nonlicensed operators properly operated 4
the AFH (ICS)- flow control valves. An operator applied excessive force with a valve wrench to close an AFW (ICS) flow
, control valve. He did so because he had not accurately i
determined the position of the valve while attempting to shut it i
completely. As a result of his actions, the valve was damaged, i
i 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 air-operated valves, and the i use of available and backup indications to determine valve
- positions. These weaknesses suggest areas where hands-on I
training rather than walk-through or talk-through training may i
l' be necessary. 9
?
f DISTRICT RESPONSE i
! A specific policy preventing'the use of the valve wrenches-on '
gear-drive valve operators has been established. Formal training.on
[ this policy and on the proper operation of such valves has been provided to the operators. -The training and qualification l
! requirements for operators have been changed to require that each
- . Individual operate certain valves,-such as the Auxillary Feedwater
- ' Control Valves, so that they are famillar with the feel and i
characteristics of the valve. Included in the training and operating {
I B-15 i
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d policies are.the require.ent to utilize the available indications of I valve position and the need to comunicate with the control room.
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FINDING - B.10 (x
- 10. While the deficiencies in SMUD's radiological control and emergency preparedness programs during the December 26, 1985 incident did not jeopardize the public health and safety due to i
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 f
The District 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 v
resolve details of procedures and hardware configuration which were identified as impediments to effective implementation. The December 26 event highlighted communications and procedure adherence issues.
Proactive steps have been taken by the District to get combined i
training between the Rancho Seco communicators and their counter parts in the counties. Site visits, both ways, have occurred and are i
now a part of the program. The Emergency Plan itself has been revised to improve and simplify 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 8-17
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Plan duties are clearly stated and pre-assigned within the operating j O crew. The Shift Supervisor / Emergency Coordinators role 'is to provide a
- l. the overview and direction to the crew. He is not to perform as a !
' control room panel operator.
j A longer term project is underway which intends to simplify the Emergency Plan in terms of making it more " user friendly". A major j benefit is expected which will be its ability to successfully f mitigate complex scenarios which involve multiple casualties such as, radiation release coincidence with personnel injury and a fire, j Improvements in the Radiological Controls, as observed in the i
December 26 event, are addressed in a following finding.
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f ~'s FINDING - B.11 1
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- 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. SMUO's intent to
+ . ,
4 satisfy II.E.1.2 with this alternate design was not made clear k
to the NRC staff, was not approved by the staff, and may not
! have compIIed 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
, installed at Rancho Seco. i j DISTRICT RESPONSE t
The AFW/EFIC (II.E.1.2) scope and schedule changes had been provided to the NRC staff.
3 The NRC issued Safety Evaluation Reports in January and September 1982, assuming EFIC installation. On October 22, 1982, the Olstrict l
Indicated that it would install interim safety grade AFW ,
modifications and that EFIC was separate and beyond the AFH upgrade requirements of NUREG-0737. The District indicated at that time that -
EFIC would be installed by Cycle 7. This schedule was confirmed by l the Olstrict on December 14, 1982, at which time the Cycle 7 outage I was expected to occur in the Fall of 1984. The schedule for the
, s i
8-19 i
interim safety grade modifications was specified in a conformatory tO order dated March 14, 1983.
On April 28, 1983, the District submitted a revised AFW system
- description describing the interim AFW upgrades. NRR confirmed their understanding in an SER on the status of the AFW system dated
! September 26, 1983.
i Then in a series of living schedule submittals, the District informed the NRC that the EFIC installation was scheduled in two phases (Cycle 8 and Cycle 9).
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- Note
- Describe the reasons / phases.
I Need to explain what took place between September 1983 and October 1985.
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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.
j As part of this restart action plan, the District has now determined i
i that the entire EFIC installation will be completed during the cycle 8 refueling outage.
k B-20
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l FINDING - 8.12 V
- 12. Although the RCS temperature dropped 180*F in 26 minutes, it would have had to racidly drop another 215*F (i.e., to an RCS temperature of about 170*F) while pressure was maintained at approximately 1400 psig, in order to seriously threaten reactor vessel integrity.
DISTRICT RESPONSE The District agrees with this finding, based on calculations provided by B&W. In its evaluation for the District dated February 1986, B&W l also calculated the effect of the December 26, 1985, event on the reactor vessel as a function of the number of cooldowns consumed in 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 event concluded that the event consumed "0.3 cycles" of the 240 designed. To date, all over cooling and abnormal events in combination, have consumed five cycles based upon NSS fatigue analysis. Normal cycles have totaled less than 35, leaving 200 heatup/ cool down cycles available.
This number provides sufficient margin for achieving the balanc' of the plant design life. l The Electric Power Research Institute (EPRI) applied a new nonmandatory ASME Code Section XI, Appendix XX, to the December 26 event and its effects upon the Rancho Seco Reactor Vessel.
B-21 l 3
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] The ASME evaluation procedure allows demonstration of adequate i
structural integrity of the reactor vessel beltline without doing
! further integrity analyses as long as the reactor coolant pressure I
has not exceeded design pressure (2500 psi) and T - RT has c NDTS
! not been less than 55'F during the transient. Calculations show that
, the Rancho Seco December 26, 1985 transient met these criteria.
i The reactor vessel calculations demonstrate, that the Rancho Seco reactor vessel beltline region has adequate structural integrity for i return to service without further evaluation.
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, I FINDING - B.13 l
- 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 lies in the fact that under alternate scenarios, 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.
t The District's Plant Performance and Management Improvement Program specifically addresses investigations which are of a retrospective nature and focus on preventing trips and tnereby 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
, e and temperature limits and begin to develop the characteristics of a serious or challenging event.
' l 8-23 l
FINDING - B.14 (s
- 14. It is not clear that the overcooling transient was within the
- . Final Safety Analysis Reocrt (FSAR) analysis of the Rancho Seco plant. Although PTS has been addressed generically, the FSAR accident analysis for Rarcho Seco'does not address this issue.
The most comparable analysis in the FSAR is for the cooldown due to a main steam line breik. However, this analysis included
)
i only 100 seconds of the transient. In addition, the Rancho Seco 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.
j DISTRICT RESPONSE The Rancho Seco original FSAR description of the main steam line break (MSLB) consisted of two analyses:
i 4
MSLB without ICS or operator actions The MSLB analysis with ICS actions is conservative with respect to )
maximizing offsite doses. The analysis assumes 1% failed fuel with l
l the technical specification steam generator tube leakage. The ICS
/ actions are assumed to occur to maximize the cooldown time to decay V)
B-24 j
-,----r - - --- - - e , -,,,-en- ~ , - - - - - - - - - -,.
heat removal system operation thus maximizing the releases via the
(
\ intact steam generator.
The MSLB analysis without ICS or operator action is conservative with s
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 MSLB inside i
the reactor building was raised. This concern was addressed by installation of automatic feedwater isolation, performed by the Main Steam Failure Logic (MSFL). The MSFL is independent of the ICS, consists of redundant actuation channels and is battery backed, however, it is not safety grade. Reactor Building Containment j integrity is not required for the MSLB, as the worst case for dose s/ considerations is the MSLB outside th'e reactor building.
The above design basis was clear in the original FSAR. The analysis for MSLB without ICS, or operator action, was contained in a response to an NRC question. During the compilation of the Updated Safety !
l 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 District is l
currently revising the description of the MSLB analyses in the USAR for clarity. This clarification will be included in the USAR update scheduled for submittal in July, 1986.
O B-25 i
I l
The cooldown rate of the MSLB analysis, without ICS or operator
\ action, bounded that of the December 26, 1985 event. The cooldown rate of the analysis was such that high pressure injection (HPI) was initiated in 23 seconds, followed by core flood tank (CFT) injection 47 seconds after event initiation. During the December 26 event, HPI/SFAS initiation occurred after 3 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 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 through the auxiliary 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 j Report," June 1983, is referenced and described, Numerous PTS events are evaluated in BAW 1791 including events similar to the December 26, 1985 event. BAW 1791 is discussed further in Section 8.26 of this response. l
\
/ l With the PTS rule making in December 1985, the NRC established screening criteria for PTS. The District's response to the screening B-26 l
I
l criteria has been' submitted, and will be incorporated into the USAR in the 1986 update scheduled for July 1987 submittal.
I I
l l
l l
l B-27
l FINDING - 8.15 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 mitigate 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 J operators. While the staff had raised these issues on a number of occasions over the pas't 6 to 8 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
) \s specific findings associated with these precursors include:
(specific responses follow.)
DISTRICT 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.
4 B-28
i FINDING B.15.a t
- 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 i reviews of issues associated with the NNI and ICS.
DISTRICT RESPONSE The District is currently reviewing the ICS and its power supplies.
j The reviews being performed and the modifications identified are described in section 4 of the Action Plan.
i 1
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(
! B-29
..-. . ~ . . - . . . . _ . , _ _ _ _ _ . . _ _ . _ . _ . . . . _ . _ _ . , . _ _ _ . . _ _ _ . _ . . . . . , _ , _ _ . _ _ . . _ _ , _ _ . _ . . . _ . .
1 l
FINDING B.15.b
- 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 District is reviewing the ICS as identified in section 4 of the l Action Plan. A procedure for the loss of ICS has been generated.
i 1
a 1
f 1
J d
l B-30 l
l
-,--e
h FINDING B.15.c i \
< c. In March 1979, B&W issued a report (BAW-1564) in which they l
l analyzed the reliability of the ICS. Although the B&W analysis
- noted a number of changes that appeared to be warranted in the i
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 l
limited scope and did not appear-to meet the requirements of the i
original Order. The NRC staff concluded that no immediate changes were required at Rancho Seco as a result of the 8&W
- analysis. The long-term issues associated with the B&W report l were to be considered in Unresolved Safety Issue (USI) A-47,
" Safety Implications of Control Systems." 1 4
- DISTRICT RESPONSE The ICS reliability study, BAW-1564, was litigated as to its completeness and adequacy before the' Atomic Safety and Licensing Board which issued a decision (LBP-81-12) dated May 15,-1981. The i
section of that decision pertaining to the ICS is provided below.
FINDINGS OF FACT 4
II.
A. Integrated Control System
- 18. Board Question H-C 16:
8-31 P = w ww w -ei- '
,y - -
wrg - ~-
1 t
k Is the failure mode and effects analysis for the j Rancho Seco integrated control system complete and
.1 adequate?
1 One of the long-term actions directed by-the l Commission in its Order of May 7, 1979, was that
"[t]he licensee will submit a failure mode and effects 1
analysis of the Integrated Control System to the NRC 1
Staff as soon as practicable." 44 Red. Reg. at 27779 (1979). Such an analysis was performed by B&W for Licensee as part of B&W's study of the reliability of
, the integrated control system ("ICS"). The results of B&W's reliability study are contained in B&W Report BAH 1564, " Integrated control System Reliability i
q Analysis." CEC Ex. 3.
- 19. In order to assess the completeness and adequacy of B&W's analysis, it is important first to understand i
the Rancho Seco ICS and the Staff's concerns regarding it. The ICS is an automatic control system whose t
basic function is to continuously match the unit's j power generation to its load demand. The ICS does-this by coordinating the rate of steam generation and the steam flow to the turbine. NRC Staff Testimony of Dale F. Thatcher Relative to the Integrated Control
- System.(Board Question 16), following Tr.1163
(" Thatcher ICS Testimony"), at 2. .
1 B-32 l
,_. _ _. _. . . - . _ _ _ _ _ . _ , . . , _ - . _ . _ . . _ . . _ ~ , _ . . _ _ _ _ _ _ _ . . . . . . _ . . _ _ . . _ . . _ . . . _ _ . , -
I s 20. During normal operations, the ICS provides proper !
As coordination of the reactor, steam generator, feedwater control, and turbine. Proper coordination consists of produc2ng the best load response to unit load demand within the limitations and capabilities of the plant equipment. Id. at 3.
I
- 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 O' controls. Id. at 3. The ICS can maintain a constant average reactor coolant temperature at power levels between 15% and 100% 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%
l to 15% and maintain that level should the turbine trip l 1
1 I
B-33 1
without calling upon the reactor's protective systems V (Karrasch-Jones testimony at 10), although presently j 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 condit kns 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 auxiliary feedwater is available. Tr. 1105, 1118, 1119.
O 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 secondary 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; Xarrasch-Jones-Testimony at 7. It appears that, in the days shortly :
after the THI-2 accident, the Staff was concerned that B-34
l
! I 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 malfunction could prevent auxiliary feedwater (AFW) i from being supplied during a loss-of-main-feedwater j transient or could cause such a transient. Id.; Tr.
! 1270-72.
! 23. The first concern was addressed on a short-term basis 3
i in tiie Commission Order of May 7,1979, by requiring i
Licensee to "[d] develop and implement operating 1
l procedures for initiating and controlling auxiliary
! feedwater independent of Integrated Control System
- control." 44 Fed. Reg. at.27779 (1979). The adequacy l of Licensee's compliance with this aspect of the May J
I 7, 1979 Order was established by the Staff by visiting I
the site and conducting examinations of the operators J to verify the adequacy of their training. This evaluation included a walk-through of some of the procedural aspects of manually controlling AFW j independent of the ICS and a review of plant diagrams i
to verify that the valves that.would be utilized for i .
l AFW flow control were indeed independent of the ICS.
Thatcher ICS Testimony at 4, 5; Tr. 1386, 3730, 3731;.
Staff Evaluation at 13. l l
[ B-35
1
- 24. A permanent solution to the first concern has been
- s. provided by Licensee's safety-grade AFH control system-independent of the ICS. This modification will completely remove the operation of the AFW system from the ICS. Thatcher ICS Testimony at 5; Tr. 1273.
l l
- 25. It was the second concern relating to the ICS that led .
the Staff to ask that a failure mode and effects analysis ("FMEA") of the ICS be performed. Since the Staff was interested in the potential role of the ICS 4
as the instigator of a transient, it sought to have an analysis made of the reliability of the ICS and the
]
effects of failures of that system on the plant's 4
operation. Tr. 648, 937-39; Tr. 1270-73. A FMEA is a i s 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 1 Testimony at 6. Accordingly, an ICS FMEA was one of j t
the long-term actions directed by the Commission in i its Order of May 7, 1979. As a long-term action it was not a condition of restart.
l I
B-36
- 1 i
- 26. B&W performed the FMEA as part of its reliability LdA) 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 i
Rancho Seco plant was chosen specifically as a representative design for all th'e B&W units for,the analysis. The analysis was complemented with an evaluation of. field dita from all B&W operating plants l
and'a computer simulation to confirm the effects of various ICS failures on associated equipment.
Karrasch-Jones testimony at 11; Staff Ex. 5 at 3. The analysis was made a part of our record as CEC Exhibit I
- 3. " Integrated Control System Reliability Analysis,"
BAW--1564, August 1979, as was a review by Oak Ridge National Laboratory of the analysis (Board Exhibit l 1). Also a part of the record is Staff Exhibit 5, the Staff review of both reports.
i 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 detail, (both of these specific to Rancho Seco), and a review of operating experience from all B&W operating plants. Board Ex. I at 5.
.i l
i B-37
~.
i i
- 28. The overall conclusion of the FMEA was that the 4
reactor core remains protected throughout any of the j 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 i
4 caused and, accordingly, its net 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 j safety system functioning and that the operating i
experience demonstrates that the,ICS is a reliable system with regard to preventing plant upsets.
Karrasch-Jones Testimony at 11-12.
i
- 29. The ORNL Review concluded that although the ICS and relateo control systems contain areas which can be
- potentially improved, the ICS itself has proven to 2
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 i
I 8-38
l l
1 l
malfunctions resulted in a reactor trip (approximately
. 6 or 162). In its review, the ORNL concluded that'the l ICS is a "significant asset to plant safety and availability." Board Ex. I at 11. ,
- 30. While agreeing with B&W's findings and conclusions and I
with the recommendations made by B&W for further 1 improvements in areas relating to the ICS, the ORNL Review pointed out a number of perceived deficiencies in B&H's approach to the FMEA portion of the reliability analysis. Tr. 1706-07, 1774. Board Ex. 1
, passim. The main criticism leveled at the FMEA by j ORNL was that the scope of the FMEA was too limited,
- leading to results having only limited value. Board i
Ex. 1 at 4. The scope limitations identified by ORNL 4
were: (1) not considering the interactions between plant safety and nonsafety systems such as ICS; (2) i not including analysis of failures of plant systems
, external to the ICS; (3) not considering multiple
~ '
system failures; and (4) utilization of functional versus component diagrams as the building blocks in the analysis. Board Ex. I at 3, 4, and 6 through 8.
- 31. It was, indeed, critical language from Board Exhibit 1 that formed the basis for this Board's inclusion of BQHC 16 in this hearing. In particular, such s a ements as.
B-39
...the B&W analysis is more notable for what it does not include than for what it does include."
1 and
...Because of this limited scope, the results are of limited value."
(Board Ex. I at 3 and 4) i
- would surely give one pause if taken out of context.
] We note, however, the following points about each of the four numbered limitations of scope set forth above:
l Point (1): Interactions between safety and l
1 nonsafety systems such as ICS were not
- considered. That is true, but such analysis was
, not specifically required by the NRC's May 7, a
j 1979,' order. A study of such actions is underway j for all plants as a part of the Staff's j " Integrated Reliability Evaluation Program" 4
l, (IREP) which has as one of its objectives to
{ identify the risk significance of systems l Interactions originating in the ICS of B&W i
plants. Thatcher ICS Testimony at 8.
Point (2): Failures in systems external to the ICS were not included. This is beyond the scope i
of the May 7 Order. Actually, the B&W analysis )
i did include some such failures in that it
\
8-40 l
l
A included failures in the inputs to and outputs
/
U} from the ICS. Tr. 681-83, 1083-86.
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-j is usually used to determine whether a single failure can prevent operation of a safety
,t
] system. Id. The ICS has not been required to j meet the single failure criterton and was not
! previously analyzed; such analy' sis 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): Functional 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 l 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 I
by examining specific component failures. Board
'Ex. I at 6. It is also possible that (if the failure rates of specific components were known) i one might estimate the probabilities of various 1
\
l B-41
,- - - - - - , - - - - - , , , - - , , - , _ - . - ,- ,,, - _ . , , - - - , - , , , , , . , - . , - , , . - , - - - - - ~ - -
l 4
modes of failure by that method. Tr. 1086.
v However, by taking the approach which they took l
the B&W analysts clearly met the requirements put upon them. Further, it is not clear to the Board l I
that a component-based analysis and estimated failure rates would give a clearer picture of reliability than the " actual history" approach which B&H supplied in addition to the FMEA. We j
think, in fact, that the reverse is true.
- 32. We note that the first conclusion of the B&W analysis was that: !
'l
)
- 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 l
l Non-Nuclear Instrumentation power supplies that i
Initiated the incident. 3717-18, a
- 33. In other areas identified by the study, Licensee is considering changes to increase the reliability of the i
reactor coolant flow-input signal to the ICS (Tr.
j 3703-04), and has developed procedures to improve the i
- 8-42
l
" tuning" of the ICS to the balance of the plant, V having trained operators further in ICS control. Tr.
3704-05. .
l 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 4
it causes and (2) the system is generally superior to manual or fragmented control schemes. Board Ex. 1 at 15.
1 .
I i 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 i
the FMEA was adequate and complete for its purpose.
t We note that it raised other issues whose resolution i
j would be expected to yield an even more reliable and safer plant (para. 33 supra), and that those.lssues j
are being acted upon. Although the need to perform a broader study of the B&W control system and its role j in the initiation and the mitigation of transients has been identified and it will be carried out in the l IREP, we see no reason to believe that the Rancho Seco B-43
I plant would present a hazard to public health and safety during the ongoing investigations and upgrading.
i 1
j l
i ,
i I
I i
i B-44 i
1
}. I I
d B&W recommends that the input signals from the NI/RPS system to the ICS - specifically the RC flow signal - be reviewed for j possible changes to enhance reliability and safety.
4 ,
DISTRICT RESPONSE This recommendation is being implemented as described in section 4 of the Action Plan.
4 RECOMMENDATION 3 4
B&W recommends that the ICS/ BOP system tuning, particularly i
j feedwater condensate systems and the ICS controls be reviewed for possible changes to enhance reliability and safety.
f i DISTRICT RESPONSE i
i The District had reviewed the ICS/ BOP system tuning and had performed ICS tuning prior to the December 26, 1985 event. ICS tuning will
) also be performed during restart.
i l RECOMMENDATION 4 l .
B&W recommends that main feedwater pump turbine drive minimum speed control be reviewed for possible changes to prevent loss .
j of main feedwater or indication of main feedwater to enhance
! reliability and safety.
I i
t B-45 1
i DISTRICT RESPONSE l
i The District is currently evaluating this concern as identified in
- j. section 4 of the Action Plan.
RECOMMENDATION 5 1
1 i
! B&W recommends that a means to prevent or mitigate the 1
i consequences of a stuck-open main feedwater startup valve be reviewed to enhance reliability and safety.
i -
i
,i DISTRICT RESPONSE l
l i
- The District is currently evaluating this recommendation per section 4 of the Action Plan. The Installation of EFIC will resolve this issue.
j i
j RECOMMENDATION 6 l
}
i j B&W recommends that a means to prevent or mitigate the i
! consequences of a stuck-open turbine bypass valve be reviewed to l
l enhance reliability and safety.
1 l
! DISTRICT RESPONSE i .
The District is currently evaluating this recommendation per section l
i 4 of the Action Plan.
i j B-46 i
.-..---..n - - - - , - , - , - ~ , - - - . - - , - . , . , - - . - , , . - . - , . . . - _ - - - . , , - . . - - - - . , - - - - - - . - - , . , ~ . . - - , - , . , .
FINDING B.15.d a
- d. As a result of the loss 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 raised significant concerns about the consequences of a loss of power to instrumentation and control systems, SMUD concluded 1
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 wtth 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 SMUD'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. !
i f
4 I
i B-47 i
4 i DISTRICT RESPONSE
\
j The District is revisiting Bulletin 79-27 as a part of the precursor review program to ensure that the concerns identified have been 4
adequately addressed in a broad scope fashion. In addition, reviews of, and modifications to, the ICS are identified in section 4 of the Action Plan.
1 FINDING B.15.e
- 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 l ('~'s response to this issue did not include the other valves at Rancho Seco that repositioned on loss of ICS power (i.e., they 1
confined it to the narrow issue associated with the ADVs). In j addition, SMUD deferred this narrow issue to installation of the EFIC system, which to date has not been installed at Rancho l Seco. The NRC found this response to be acceptable.
1 i
j DISTRICT RESPONSE 1
{ Following the December 26, 1985 overcooling event, the District j performed modifications to the controls of the atmospheric dump j valves, turbine bypass valves (TBV), and the auxillary feedwater (AFW) flow control valves. These modifications are discussed in
! section 4 of the Action Plan.
! l B-48 4
1 1
t i .
FINDING 8.15.f i
i i
< f. Because of concerns about the transient response of B&W-designed reactors and the role of ICS as an initiator of such transients,'
i NRC conducted an extensive study and made 22 recommendations in l- NUREG-0667. However, it does not appear that these i
recommendations were sent to SMUD for action or that the recommendations that are relevant to the December 26, 1985 l incident were implemented at Rancho Seco, t
DISTRICT RESPONSE l t
l l The District has implemented many of the NUREG 0667 recommendations.
l The balance of the recommendations are currently being evaluated by i the District and/or the B&W Owners Group.
o I
f I
1 1
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4
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i B-49
t FINDING B.15.g
%O
- g. The March 29, 1984 partial loss of NNI power at Rancho Seco again demonstrated that the failure of nonsafety-related a
, 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 9
primary system. Despite the fact that this event occurred J
nearly 2 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 Rancno Seco, the Team again i sees strong evidence of a narrow focus on the incidents V 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 The District has implemented modifications and is 'In the process of implementing additional modifications per section 4 of the Action l Plan, to ensure that the plant will go to a known statt following a 1
loss of ICS or NNI power. .
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B-50 i
FINDING B.15.h
\O
- h. While the scope of the analysis performed under USI A-47 is -
i broad, it appears that to date the actual study includes only those events with the potential to produce consequences outside the desiga 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 j }
not adequately considered. Therefore, it appears that the 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 December 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 I
that was studied. The results are not directly applicable to most other B&W-designed plants such as Rancho Seco because of I the differences in the design of the ICS.
DISTRICT RESPONSE 1
The resolution of USI A-47 is a NRC staff action. However, the O g District has implemented a program, as described in the Action Plan,
,i B-51
1 .
- r j to identify events that challenge the safety systems or the operators l and take appropriate corrective actions. In addition, the B&W Owners '
I
] Group has embarked on the Stop-Trip Program, which is aimed at l reducing the frequency and severity of transients at B&W plants. A '
i portion of this program includes an extensive evaluation of the ICS l as a contributor to plant transients. This evaluation will take into i
i consideration the differences in ICS designs between the various B&W i
j plants.
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FINDING - B 16
- 1. It appears that the transient initiator (i.e., the loss of ICS de power) was not fully recognized by control room operators l
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 I
which requires immediate operator action: 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 t.nnunciator alarm and the loss of ICS Controller Hand / Auto Indicator lights. Coupled with the immediate "under cooling" effects (due to the Main Feedwater pumps being runback) there was no ambiguity which suggested a " Fair Power Failure." In response to the' potential for ambiguous annunciator alarms, Engineering Changes R-0517 and R-0580 l have been implemented to provide discrete alarm windows for both ICS and NNI Power failures separate from any other alarms. Furthermore,
'I as a long term effort a complete reassessment of the Annunciator Systems is to be accomplished to insure that. annunciators are provided for all important parameters and that they are unambiguous
, to the operator. This is an element of the CROR Human Factors l
[Vh upgrade.
B-53 l
I
[ FINDING - 8.17
- 2. The Annunciator Procedures Manual was not used by the operators following the "ICS or Fan Power Failure" alarm. Even if the 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 de 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 /CRDR Annunciator upgrades which are also being developed.
l B-54 i
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FINDING - B.18
('
- 3. 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.).
4 DISTRICT RESPONSE
- The B&W Owners Group is evaluating various aspects of the ICS including performance upon restoration of power. This evaluation O' will consider the results of testing performed at Davis-Besse on the ICS performance upon power restoration.
The District has installed modifications to the important ICS controlled values which provide power and controls in the control room which are independent of the ICS. This ensures that the demands from ICS during restoration of power will not cause a subsequent transient. Folicwing repowering and stabilization of the ICS, the operator can return control of each device to the ICS satisfied that the transfer will be "bumpless". Use of the controls has been incorporated.into the " Loss of ICS" procedure and training has been 4
performed. Proper functioning of these new controls, and validation of the new procedure for repowering, will be demonstrated during the
( restart. test program. Also See Section 9.7.
B-55 l .- . .-
< N FINDING - 8.19
- 4. 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 (MFW) flow recorders are affected by the ICS. During the December 26, 1985 incident, the recorder failed to a value near mid-scale when MFW flow was actually zero.
DISTRICT RESPONSE i -
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 j for information only to the operator. The Main feedwater Flow Recorders are affected by the ICS as it is necessary to combine the Startup and Main flow signals to develop the total feedwater delivered to each steam generator. The mid-scale value which i
results, following loss of ICS power, has been determined to have noi adverse effect upon operations should this event reoccur. The reason is that a loss of ICS power will cause the main feedwater pumps to B-56
-1
l' I
runback to minimum speed while the feedwater control valves go to mid-position. The runback causes an undercooling which results in a reactor trip. Since the indication of Main Feedwater Flow remains high (mid-scale position) the operator would expect an overcooling to i
result. The Emergency Operating Procedures (EOP's) direct that the associated pumps then be tripped, in this case the Main Feedwater Pumps. Since' they have already runback to minimum speed, the trip would have little effect on the transient. Section 4C.1.d provides a description of the detailed actions being implemented to enhance indications upon loss of ICS power.
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- 5. Because of a perceived sense of urgency, two-nonlicensed operators made an emergency entry into the makeup pump room without respiratory protection or adequate protective clothing, neither of which was readily available. As a result, their clothing was contaminated and they were exposed to airborne radioactivity.
DISTRICT RESPONSE Since the event, the appropriate procedures have been revised and training has been completed in response to the lessons learned.
1 Furthermore, managements policies regarding radiation protection and procedure adherence have been clearly stated to insure that all personnel understand their responsibilities. Additiona1' 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 bee'n 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.
U B-58
FINDING - 8.21
- 6. The operators did not remember a recent modification had been-made to permit the TBVs and ADVs 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 shut down panel controls were 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.
O .
B-59
FINDING - B.22
- 7. Additional staffing above that required by plan't Technical f
- 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.
f 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 O at any point by simply performing the E0P step to " trip the appropriate pumps." It was a conscious decision by the operator-to !
not perform that step. That decision process has been~ resolved by. i subsequent management policy, procedure changes and operator training. Furthermore, the District has implemented modifications !
and procedural changes which would decrease the demands upon the l operators by providing controls in the control room which are powered independently of the ICS. This would eliminate the need to dispatch i operators.oct into the plant.
+
t 8-60 l
O FINDING - B.23
- 8. Neither the operators nor the Shift Technical Advisor (STA) could identify an instance of when the STA provided engineering j expertise during the incident. However, the operators found the STA valuable as an extra person on shift to help out during the l l
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 i Functions in an independent overview role and that the operators have engineering expertise available when needed. The STA provides valuable support to the operators and the District intends to continue to support their utilization in this role. They are vital 4
members of the operating crew and can significantly enhance nuclear safety.
O
( B-61 1
FINDING - B.24 C
- 9. It appeared to the Team that SMUD personnel found the process of troubleshooting in a highly controlled, systematic, and well-docurranted 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 institutec a systematic program for analyzing the event and resolving root causes. The program was based on NUREG-1154 Appendix.8, 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 comparison of the District's program with NUREG-ll54 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 the District's opinion, did not constitute (O substantive changes affecting the outcome of the troubleshooting or-the effectiveness of the program.
B-62
i'
- 1 i l
, The District had a highly controlled , systematic, and well documented troubleshooting program in place prior to, and following the i
l- December 26, 1985 event. :
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FINDING - B.25
- 10. .Throughout the Team's review of the December 26, 1985 incident, SMUD personnel had considerable difficulty providing information in the detail that the Team requested. Thus, SMUD personnel repeatedly summarized data, analyses, and p1(ns 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 District's investigations following the December 26,1985 event were bree.d in scope and exceeded that identified by the IIT. As the IIT increased their knowledge of the plant and the event, they expanded their areas of interest. Often the District already had an investigation underway in the area of question. (This may have led the IIT to perceive that the District was not providing information in the detail requested.) When requested the detailed information was provided, as stated in Finding 10.
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.
O B-64
FINDING - B.26
\
- 11. 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 3 years. Thus, it would appear that this analysis predicts that events comparable to the December 26, 1985 incident wculd o'ccur 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 BAW 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.
O B-65 l
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f APPENDIX C CROSS REFERENCE ACTION PLAN i
TO i
1 j NRC OPEN ITEMS I
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APPENDIX C l CROSS REFERENCE TO NRC OPEN ITEMS A. Areas of Concern Relating to December 26, 1985 Overcooling Event (NRR and Region V) :
A.) Loss of DC Power to ICS/NNI I -
Root Cause RJR 86-75 See Root Cause Report 85-41 (VII.1)
Control Room Instrumentation Failure B.16, B.17, B.19
, (e.g., Main feedwater Flowmeter i 501 readout)
FMEA 4C.I.e i
- Power Supply Honitors 4C.I.c i - Integrity of Electrical Terminations 4C.I.a.1.5, 4C.2.a.i.5, 4C.13.1.6 A.2 Plant Response on Loss of ICS/NNI f -
ADV and TBV 4C.I.b.1.2
) - Startup and MFH Control Valves, RJR 86-75 -
HFP Speed (V.I.4.b)
I - Control Independent of ICS for AFH 4C.I.b.1.1 i Control Valves ADV and'TBV t i
Secondary Steam System Valve 4C.3.d RJR 86-082 Isolation Halkdown Check Capability (VI.4.1)
'of Isolation from Control Room Response to Return of ICS Power B.7 and B.18 i C-1 i ,
_ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ , _ - _ _ _ - e
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l APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS A.3 Hakeup/HPI Pump Failure ,
i -
Root Cause RJR 86-75 See Root Cause Report 85-41 (VII.3)
- Interlocks on Water Supply Sources 4C.12 I -
Makeup Pump Repair 4C.13.1.3 A.4 Overcooling Effects and Reactor Vessel and Steam Generators 6
Analysis of Internal Transient on RJR 86-75 i Vessel and Steam Generators (V.6) ,
- License Amendment to Clarify Include in Restart Report Cool Down Rate ,
Technical Basis for PTS Guidelines include in Restart Report Potential for Core Lift Include in Restart Report
! A.5 Radiation Monitor Damage Root Cause -
RJR 86-75 (V.5) l -
Effects of Containment Isolation Include in Restart Report i
l - Systems required following ESFAS Include in Restart Report i
l i
1 C-2 I
I l
i APPENDIX C l CROSS REFERENCE TO NRC OPEN 11[MS i
A.6 Flooding of Hain Steam Headers -
l - Evaluate Steam Header Supports 4C.3.d.1.3 Engr. Analysis Complete. Lines OK.
! Include in Restart Report ,
i t i A.7 Plant Maintenance and Testing
- Evaluate Maintenance Program 4C.13.1.8 i 4C.13.2.1 l 4C.13.2.4 Manual Valve Maintenance 4C.13.1.9 Preventive Maintenance and Other 4C.13.2.4 Deficiencies l
- Troubleshooting for Root Cause Include in Restart Report Determinatton Reconsnenda t ion 21.0007.H Repair Damaged Valves 4C.13.1.10 Work Complete !
Systems Test Program 40 I
4 7
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C-3
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1
APPENDIX C I CROSS REFERENCE TO NRC OPEN ITEMS A.8 Training and Operator Performance Adequacy of Licensed Operator and 4B.3 RJR 86-75 Non-Lit.ensed Operator Training (V.9) i o ICS Off Normal Operations 4C.I.a.1 o Operation on Makeup System 48.3.2 ,
, o AFH Throttling 48.3.2 o AFH Pump Trip Criteria 4B.3.2
- o Ewrgency Plan 4B.3.4, 48.8 i
o Communications 48.8 o ADV, TBV Operation 48.3.2 o Differences Between B&W 48.3.2 RJR 86-75
- Simulator and Rancho Seco Plant .
(V.9.8) o Operation of Manual Valves 48.3.2.1.c o Plant Modifications, Procedural 4B.3.2 Changes and Additions 1 - Minimum Staffing Requirements B.22 4
- Security / Safety Interfaces RJR 86-75 (V.7) l I
4 C-4 i
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l.
l l APPENDIX C i CROSS REFERENCE TO NRC OPEN ITEMS l !
- A.9 Normal and Emergency Procodures ,
I Event Related Procedure (Loss of 48.4.1 RJR 86-75 !
I ICS Power) (V.9.4) !
Adequacy of ATOG for PTS B.6
- Adequacy of Health Physics and 4B.3.3 . .
Energency Procedures B.10, and 8.20 ,
Adequacy of Annunciator Procedures B.17 Hanual j i - Hethodology for Ensuring Changes Management Issue Section.
Properly Reflected in Procedures '
l i
A.10 Human Engineering Deficiencies !
i
- Breaker Position Indication 48.9.1.1.e and B.8
- -- Valve Position Indication on TBV, 48.9.1.1.b ADV and AFH flow Control Valves B.7 and 8.9 q j -
"ICS" and'" fan failure" Alarm 4C.1.d. 4C.2.c, ,
3 and B.16
- ~
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! C-5
! I l l APPENDIX C CROSS REFERENCE TO NRC OPEN Il[MS l A.11 Retrospective Issues
, - Adequacy of FSAR Analysis B.14
- EFIC System B.11 o Installation Delay o Current Schedule
! o NUREG-0737 II El.2 Justification
- Modifications and Improvements of B.15 Other Facilities t
i - Reliability of ICS 8.15 i o Bulletin 79-27 I o CR-3 Feb/80 Loss of NNI Power 1
o NUREG-0667 '
4 o Partlai Loss of NNI at Rancho Seco March /1984 l
- Probability of PTS Events o BAH-1791 B.26
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3
I i APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS t
l ,
B. Region V Additional Items l .B.1 Post Accident Sampilng System 4C.6 i System Modifications t
- Procedures and Training j i
Testing I B.2 Control Room /ISC Emergency HVAC System '4C.7, 48.4.1.3 i
Adequacy of Design and Installation Modifications
- Loading A Train on Diesel Generator
. B.3 125V DC Station Batteries J
- Adequacy of Batterles Include in Restart Report l
-- Replacement 4C.13.1.4 i
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$PPENUlX C CROSS REFEkcNCE TO NRC EsPLN ITEHS 8.4 Radioactive Liquid Effluent Releases 48.6 and 48.7 3
Offsite Contamination Technical Specification Deficiencies
- Long Term Resolution
, B.5 Emergency Plan 48.8
- Heterology Program Improvements 1
j - Training 48.3.4 I
i -
Procedures and Dose Assessment 4C.9 Addresses Reactor Building
- Purge Only i
C. Licensing Areas (NRR) NOT REQUIRED FOR RESTART, PER NRC I
, C.I Regulatory Guide 1.97 .
1 Living Schedule Implementation INSTALL BULK IN CYCLE 8 REFUELING OUTAGE, SOME IN CY9R.O.
IN RESPONSE TO NRC RAI i
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1 4
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j APPENDIX C j CROSS REFERENCE TO NRC OPEN ITEMS J
C.2 Control Room Design Review (DCRDR) l
- Living Schedule Implementation 48.9.2.3 STUDY COMPLE1E END Of 1986 ,
j MODS INSTALLED CY9R.O.
C.3 SPDS.
- Living Schedule Upgrade to Safety Grade HILL BE CLASS 1 H/E Of CRTs, SEISMIC CRTS CY9R.O.
Isolation Devices I&C HILL ADDRESS (RECENT ISSUE) i
! UNDER REVIEH BY HUMAN FACTORS Modifications (format) ,.
3 C.4 Spent fuel Pool Cooling System i
- Upgrade of Cooling Return Line' CY8R.O. (LIVING SCHEDULE ITEM)
C.S Inadequate Core Cooling Instrumentation
! - Implementation -
CY8R.O. (LIVING SCilEDULE ITEM) i
, C.6 Class IE Electrical System i
I -
TBI Diesel Generator Qualification CY8R.O. (TESTING THIS SUMMER) ;
LIVING SCHEDULE ITEM I :
- Complete Class IE Electrical System CY8R.O. (LIVING SCHEDULE 11EM) in NSEB .
t i
C-9 l ;
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APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS 1 2
3 l C.7 Low Temperature Overpressure Protection Install Bypass Valve in Makeup CY9R.O. (LIVING SCllEDULE ITEM)
C.8 Security Hodifications !
i NSEB, Diesel Generator Building . CONSIDERED BY NRC TO BE TIED TO DGB/NSEB OPERABILITY, L.S. INDICAIES APRIL 1986 DATE.
C.9 Relief and Safety Valve Test Install Pressurizer Safety Valve CY8R.O. (LIVING SCilEDULE ITEM)
Supports C.10 Reactor Coolant Pump Seal Damage Automatic Initiation of Seal CY9R.O.(LIVING SCllEDULE ITEH) l Injection on Loss of Offsite Power 4
l C.11 Appendix R i -
Complete Alternate Shutdown CY8R.O. (LIVING SCHEDULE ITEM)
(Additional Isolation Switches) l - Automatic fire Suppression NSEB CY8R.O. (LIVING SCllEDULE ITEM)
(Areas B1 & B2)
L - Circuit Separation of NSEB HVAC CY8R.O. (LIVING SCHEDULE ITEM) i Cable Reroute llPI Train A CY8R.O. (LIVING SCilEDULE ITEM)
C-10
i i
4 APPENDIX C
- CROSS REFERENCE TO NRC OPEN ITEMS C.12 Seismic Assessment of foothills fault NRC ACTION REQUIRED System .
1
- C.13 Compilance With 10 CFR 50.46 (NUREG-0737 B& HOG EFFORT, DUE Jul.Y 1986 (CCL ITEM) i II.K.3.31)
,) C.14 Diesel-Generator Reliability Technical TO BE SUBHITTED lHIS SUHHER
- Specifications (GL 84-15) t -
C.15 Revised Technical Specifications for NRC ACTION REQUIRED, SUBMITTED TEB. 1986
, Containment Isolation Valves i
j i C.16 10 CFR 50.62 Requirements for Reduction NRC ACTION REQUIRED ON B&WOG DESIGN SUBMITTAL, l SCHEDULED FOR CY9R.O., ON LIVING SCHEDULE I- of Risk from ATHS Events i
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1
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1
. . . - _. _. -. ._ ._. . .___ _ . . _ . _ . . . _ ~ . . _ . - _ . - _ _ . - _ _ _ _ _
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4 1
)
i APPENDIX D' i
J j THE DISTRICT'S i i ASSESSMENT AND COMPARISION i
0F '
i ,
i' TEST PROGRAMS .
'l
- 1 i
RANCHO SECO (ORIGINAL STARTUP)
I RANCHO SECO (RESTART) )
- DAVIS BESSE (RESTART) l 1
4 TMI 1 (RESTART) i l
i I
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--- - . _-.. - . .- _-- _~. - - - . .- - - - . _ . -
J '
APPENDIX D TEST PROGRAM COMPARISON a
l SYSTEM RANCHO SECO (ORIGINAL STARTUP) RANCHO SECO (RESTART) , DAVIS BESSE (RESTART) THI 1 (RESTART) i Normal Reactor Building Purge Funct- Reactor Building Purge Fan Post HVAC ional Test Maintenance Test i Inplace Testing of Charcoal and Auxiliary Building / Spent Fuel Spent Fuel Pool Normal Ventila-2 HEPA Filter Banks Building hEPA Filter / Charcoal tion System 18 Month Test Efficiency Test Functional Test of Reactor Building Normal Cooling Verify integrity of carbon and HEPA filter banks for Battery ,
Butiding Control Room Normal Ventilation Test Control Emergency Control Room Filter- Evaluate effectiveness of noise Emergency Ventitlation Monthly ing System Balance reduction mods Test. CR Emergency Ventilation Room 18 Month Special Test. CR Essential HVAC Verify methyl iodide removal Emergency Ventilation Refueling efficiency of carbon, integrity Test. Chlorine Detection of carbon and HEPA filter banks. System 18 Month Test. CR system integrity, system func- Normal Ventilation. CR Emer-tion gency Vent System (Isolation Dampers). ,
a Cooling Capacity Test a
" i t
Control Room Pressure Boundary Test
) Instrument Compressed Air System Func- Diesel Driven Air Compressor /
tional Test Dryer Startup Testing Functional Test of Backup Air Verify adequate air supply to
' Supply to Selected Pneumatic Emergency Feedwater (EFW)
- valves Control valves and turbine driven EfW pump steam control valve with loss of instrument and backup air (3% power)
Verify bottled air supply is capable of supplying air for a minimum of two hours follow-ing loss of normal and backup instrument air 4 Test N2 System - IAS Crosstie i
1 1 1: s
/
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1
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SYSTEM RANCHO SECO (ORIGINAL STARTUP) RANCHO SECO (RESTART) DAVIS BESSE (RESTART) THI 1 (RESTART)
Pipe Piping and Equipment Displace- Verify hydraulic safety snubber None Perform thermal checks of:
Supports ment Test (Scratch Gage Test) operable by inspection and func-tional testing. Those variable / constant sup-ports (springs, counterpoises and snubbers) on pipe systems which have been added or mod-
' ified subsequent to the original Plant Startup Hot Functional testing.
Those variable / constant sup-ports (springs. counterpoises and snubbers) on pipe systems which have not been modified subsequent to the original Plant Startup Hot Functional testing, but may have been affected by those which have been added or modified.
All ma,in steam heated post supports.
53 All supports inspected during na Hot Functional testing.
Rigid supports, inspected as necessary during walkdown.
Verify that the applicable system (s) piping can expand without obstruction during heatup to nornel operating conditions.
Perform system balance and establish integrity of pipe supports in the areas of operability and functional performance.
4 Inspect those piping systems and their respective supports which are considered cold; t.e., no thermal or terminal movements.
SYSTEM RANCHO SECO (ORIGINAL STARTUP) RANCHO SECO (RESTART) DAVIS BESSE (EESTART) TMI 1 (RESTART)
Main Turbine Pre-Roll Electrical None Main Generator. Main Trans- Turbine Generator Operational
- Generator / Test former and Auxiliary Trans- . Testing.
Exciter for1mers. Protective Relaying Schemes Operational Test. Record stator slot and coolant temperatures on the Turbine Generator at various power levels Waste Gas Waste Gas System Functional Verify system integrity None None ,
Test i
Waste Gas Samp1tng System Functional Test !
e i
t 1
j s
i i .
i
-~_ . _ . -. . .. . -.
O O .
SYSTEH RANCHO SECO (ORIGINAL STARTUP) RANCHO SECO (RESTART) DAVIS BESSE (RESTART) THI 1 (RESTART)
Fire Diesel Fire Protection System Diesel Fire Protection System None Protection Pump Weekly Surveillance Test Pump Weekly Surveillance Test beter Fire Protection System Electric Fire Protection System Electric
- Pump Weekly Surveillance Test Pump Weekly Surveillance Test Fire System Valve Surveillance Fire Protection System Valve Test Survet11ance Test Sprinkler System Valve Fire Protection System Flush Surveillance Test Fire Detection System Func- Non Supervisory Fire Detection tional Test Functional Test Fire Barrier Inspection Transformer Deluge Test Verify that piping modifica-tions to suppression system piping in zone 104 were in-stalled properly and the desten intent was achieved C3 Verify final installation and I, settine of CO2 systems in zones 75 through 80 PASS N/A Verify the desten modifications Operational test of PASS. "Re-made to the PASS were installed mote Shutdown. Post Accident properly and the desten intent Monitoring Instrumentation was achelved. Channel Check."
Verify the capablitty to obtain and analyze reactor coolant and Reactor Building atmosphere '
samples under accident condt-tions.
Verify instrumentation assocta-ted with PASS is functional i
-. _ _ - . . -.- . - - . - . - - .- . - - - . .. - . __ .- ~. - . - - . .. . . - . - - .
N
\
j SYSTEM RANCHO SECO (ORIGINAL STARTUP) RANCHO SECO (RESTART) DAVIS BESSE (RESTART) THI I (RESTART) 208 VAC Emergency Lighting and Power None Auxiliary Building Non Rad HVAC None Supply Transfer Test (Low Voltage Switchgear Room) 120 VAC Vital AC, Computer and Control None Capability to Cross-tie MCCs None 1 Power Supply Tests (Includes E11B and Fila (480V-A.#)
456-2 and 456-3) l
!' 480 VAC Energizing and Interlock Test None 480V Unit Substation Live Verify system function via of 480 KV Switchgear Manual Transfer Test Surveillance Procedures (13.8KV-TPO-1 V.D.1) 4160 VAC Energizing and Interlock Test None On site AC Bus sources, lined None of.480 KV Switchgear up and available (modes 5.6)
(V.D.2)
On site AC Bus sources. lined up and available and isolated (modes 1-4) 4160 VAC Systems Transfer and j3 Lockout Test Un ,
Integrated SFAS or 18 Honth Test (with T-Mod to record load of essential 480V Unit Sub- >
station V.A.2).
I' 12.4k VAC/ Startup Transformer No. 1 Trip None 13.8 KV Bus A (B) Lockou6 j 22k VAC Test Scheme
-Startup Transformer No. 2 Trip Independent Offsite AC Source Test Bus Transfer Main and Unit Auxiliary Independent Offsite AC Sources Transformers Trip Test lined up and available Plant Plant Computer Field Verifica- None Verification of Computer None Computers Lion Test Calculation Computer Input-Output verifica- TS Computer Alarms Functional tion Test Check Computer Controlled Turbine Startup Functional Test i
[
d U (%,
SYSTEM RANCHO SECD (ORIGINAL STARTUP) RANCHO SECO (RESTART) DAVIS BESSE (RESTART) TMI I (RESTART)
Steam NSS Accept.*nce Test Examine 12,000 tubes (TS 4.17)
Generators Steam Generator Secondary Hydro Verify by system walkdown and Response time testing of M31n (Includes 276-1) secondary system sampling that Turbine Stop valves. Main OTSG pressure boundary is Steam Isolation valve Response
, intact Time Test (MS-TPO-07)
OTSG Level Verification Test Primary to secondary leakage Miscellaneous Valve cold Shut- A series of plant heatups/
check down Test cooldowns were conducted after repairs to OTSGs were com-Steam Generator Remote Shutdown pleted to test effectiveness Instrumentation Calibration of repairs Steam Generator to SFRCS Refuel-ing Calibration Steam Generator Shell Thermo-couple Readings a
Turbine Bypass Valve Instrument String Calibration C Main Steam Determination of stress in Verify Main Steam Safety Valve
/:n piping during thermal and setpoints pressure transients (strain sage test) MS and RCS Piping and Equipment Displace- Verify perforinance of Throttle Response time testing of Main ment Test (scratch gage test) stop valves Turbine Stop Valves MS and RCS Steam Dump to Atmosphere and Main Steam Isolation valve Miscellaneous Valve Cold Shut-Bypass Valve Test Stake Test down Test Steam Trop Survey AW Air Voltme Tank Pressure AW Operability Test. AVV
, Instrument String Calibration Turbine Bypass Valve Instrunent i String Calibration
[% % w
\ V (
SYSTEM RANCHO SECO (ORIGINAL STARTUP) RANCHO SECO (RESTART) DAVIS BESSE (RESTART) THI 1 (RESTART)
Reactor Reactor Protective System RPS Calibration RCS Temperature Input to RPS Surveillance Procedures apply Protection Preop. Calibration and Func- a) Temperatures Transmitters Refueling Calibration for RPS Functional Test tional Test (Includes 300-2 and b) Pressure Transmitters RCS Pressure to RPS Refueling 300-4) c) RCS Flow Transmitters Calibration d) Containment Building Pressure SFRCS and RCP Honitor Inputs e) ARTS Pressure Switches Containment Pressure to RPS f) Out-of-Core Detectors Monthly Functional Test Intermediate Range Prestartup Functional Test RPS Overall Response Time Calculations Nuclear Nuclear Service Cooling Water Nuclear Service Cooling Water Component Cooling Water System N/A Service System Functional Test Pump Operability Test Refueling Test Cooling (Includes 160-1)
Water Nucicar Nuclear Service Raw Water Nuclear Service Raw Water Pump Service Water System Honthly None Listed Service System Functional Test Operability Test Test
?
N Raw Water 4
~- - - = . . - . . . - ,., , ,
, [
V \
V SYSTEM RANCHO SECO (ORIGINAL STARTUP) RANCHO SECO (RESTART) DAVIS BESSE (RESTART) THI '1 (RESTART)
Decay Heat Decay Heat and Radwaste Room Decay Heat System Pump DH/LPI Pump and Check Valves System Pumps Functional Test Operability Test Test Decay Heat Removal System Decay Heat System Valve Stroke DHR System Isolation Test Functional Test Test Exercising of ECCS Valves Qtrly RCS Isol. Check Valves Leak 4
Verify H0V stroke times accept- Test able Operational Check of Passive Valves Operational Check of PASS DH49 Forward & Reverse Flow Test DHR System Isol. Test (ST 5051.02)
DH/1A(B) Control Power Switch Emergency Sump Valves LP Injection Header Warming Refueling Interval SFAS Test ECCS Subsystems Refueling Test Decay Heat Removal E.S. Test Pump Functional Test (Includes 204-12) Decay Heat Pump Venting Venting of ECCS Piping Borated Water Storage Tank LPI and CS Leak Rate Boron Concentration Verifica- Borated Water Storage Tank 5' tion Boron co Concentration and Volume DH Valve Pit Leak Test DH Vhlve Pit Level Alarm Test Pressurizer Pressurizer Operational and Pressurizer Code Relief Valve Pressurizer Code Relief Valve
! Spray Flow Test Setpoint Setpoint
' Pressurizer Heater and Controls Pressurizer Essential Heater Pressurizer Essential Hcater Verify Pressurizer heaters Electrical Test Bank Capacity Bank Capacity assigned to one emergency bus are sufficient to stabilize Pressurizer Level Verification PORV Position Indication Pressurizer Power Relief Valve RCS Pressure Channel Calibration Check (with major mod to change setpoints)
Pressurizer Power Operated and Safely Relief Valve Honitoring System Monthly Functional PORV/SAF Relief Valve 18 Honth Channel Calibration 1 PORV Operability (this test includes testing acoustic monitor flow alarm lights and PORV isolation valve)
I
__ - _-_ ___ __ _m-
__ _ _. _ . . _ _ ..~ _ _ _ _ _ . _ . .. . _ _ _ _ _ _ _ _ _ _ _ . .- _ __ . _ _ __ ___ _.. - ~. _ _ . _ _
)
SYSTEM RANCHO SECO (ORIGINAL STARTuP) RANCHO SECO (RESTART) DAVIS BESSE (RESTART)- THI 1 (RESTART)
I Nuclear Core Power Distribution Verify core is operating within Post Refueling Testing Critical Boron Concentration Fuel design assumptions for re-Dropped Control Rod Test activity coefficients at power Sensible Heat & NI Range Overlap Power Imbalance Detector Verify boron rundown curve is ,
Correlation Test applicable to cycle 7. Adjust React 1 meter Checkout predicted curve. if appropriate, Rod Reactivity Worth using steady state NFP data All Rods Out Critical Boron Measurement Concentration Incore Detector Testing Temperature Coefficient of i
Reactivity Pseudo Control Rod Ejection Test Control Rod Group Reactivity Worth Heasurement Zero Power Physics Testing Differential Boron Worth Reactivity Coefficients at Power Safety Rod Drop Worth i
Heasurement Incore Monitoring System Post Installation Electrical Test Ejected Rod Worth Test 7 2eactivity Depletion vs Burnup MD Core Powtr Distribution l Verification j
.! Power Imba10nce Detection Correlatiot Test
" Determination of Hot Channel Factors" i
Nuclear Nuclear Instrumentation Detec- Verify calibration of Dut-of- Incore Instrument Channel Verify functioning via Instrumen- tor Cabling Test Core monitors per TS Section 4 Calibration Surveillance Procedures tation with respect to core power and imbalance Nuclear Instrumentation Preop. Verify calibration of backup ST 5091.09 Source Range Verify proper operation of the !
Calibration recorders per TS 3.5.4 Functional Test incore thennocouples at nominal power plateaus Nuclear Instrumentation Verify Calibration of Incore ST 5030.17.06 Intermediate verify that the incore thenmo-Calibration at Power Thennocouples per TS Section 4 Range Prestartup Functional couples are giving an accurate test indication of the temperature distribution in the core and
- Incore Instrument Calibration Incore Instrumentation Channel those thennocouples which are Check symmetric to one another give comparable readings 1
_ _ . . ~ . . _ _ . _ . _ . . ___ _ __ . ._. _.__. .. .__ .-
_ _ _ _ _ _ - . _ _ _ . - . _ _m . . _ _ . . . . - . . . .
\
, l i
SYSTEM RANCHO SECO (ORIGINAL STARTUP). RANCHO SECO (RESTART) DAVIS BESSE (RESTART) THI l- (RESTART) i Radiation Radiation Monitoring System Shift Radiation Montor Channel Shift Channel Check of the Monitoring Test Check Radiation Honitoring System 3
- Monthly Radiation Monitor Monthly Functional Testing of I Calibration Radiation Honitoring System Quarterly Radiation Monitor Radiation Honitoring System Effluent and Effluent Monitor-Calibration Channel Calibration ing System Test I
Borated Boric Acid Htx and Storage CBAST boron concentration Bororn Cencentration and Volume Plant procedures apply to
- Water System Functional Test vertftcation soluble poison concentration i test
! Boric Acid Pumps Quarterly Montly Functional Testing of Operability Test Radiation Honitoring System Radiation Honitoring System Channel Calibration
! BA Flow Path Monthly Valve Vertftcation Boron Injection Flow Path
, 5 Honthly Makeup Pump f c$
, Safety SFAS Preop. Calibration SFAS Monthly Calibration SFAS Monthly Test. Containment Features (Includes 300-2) Pressure Input to SFAS Calibra-Actuation tion. Containment Radiation Input to SFAS Calibration.
SWST Level Input to SFAS Calibration.
Integrated SFAS Actuation Test Refueling Interval SFAS Test Integrated SFAS or 18 Honth Integrated E.S. actuation test Test by Surveillance Procedures BCS Pressure Input to SFAS i AuktlatryButidingSFASTestof 1 Vent System Control l
i non Nuclear Reactor Non Nuclear Instrumen- Refueling Interval NNI Calibra- RCS Hot Leg tevel Heasurement TSAT Functional Test - verify Instrumen- tation Hot Operation and tion Functional Test hot operation and calibration i tation Calibration. Non Nuclear Instrumentation Preop. Calibra-i tion (Includes 201-3, 202-6 203-5, 271-4. 300-1, 300-3)
Verify that the termination and rework program did not result i in improper terminations by verifying proper relationships i between INPUTS. PROCESS PARA-j METERS AND OUTPUTS
~ - . - -. . -
\ /
SYSTEH RANCHO SECO (ORIGINAL STARTUP) RANCHO SECO (RESTART) DAVIS BESSE (RESTART) THI I (RESTART)
Emergency Diesel Generator Electrical Verify operability of Dnergency Perfona the Olesel Generator Diesel Tests Diesel Generators and Auxiliary Load Test to ensure the Generator Systems engineered safeguards system Diesel Generator Hechanical has the capacity and capability Tests to accept to accept the loads in the required time sequence Test full load rejection. Fuel oil transfer system. 92-day
' sample. Overspeed trip test Control Circuit Fire Isolation Test Test Governor Control Diesel Air Start System Refuel-ing Relief Valve Test EDG Electrical Noise Test EDG Air Start System Test I c3
^
Reactor RC Pump Flow Test - No Core Coolant Pump RC Pump Flow Test With Core .RCP Performance Honitoring RCS Flow Rate Test (requires RCP "B" Shaf t Replacewent precalibration of miscellaneous Tests. RCP Operation / Flow instrumentation) Test records various RCP parameters including flow RCP Reactor Coolant Pumps Perforsence Procedure Safety Integrated SFAS Actuation Test Integrated SFAS Test Integrated SFAS Test None Listed Features Safety Features Electrical Test SFAS Calibration Decay Heat System SFAS Test SFAS Preop. Calibration (Includes 300-2)
Core Core Flooding System Functional Full Stroke Testing of CFT Core Flood System Refueling None Listed Flood Check Valves CFS-001 & CFS-002 Test Phase 1. RCS Isolation Check Valve Leak test Phase II
& III. Core Flood and Decay Heat System Pressure Isolation Check Valves 1
1
( ..
SYSTEM RANCHO SECO (DRIGINAL STARTUP) RANCHO SECO (RESTART) DAVIS BESSE (RESTART) TMI 1 (RESTART) f
-i
! Core Core Flood Tank Check Valve Boron Concentration (Core Flood ,
- Flood Leak Test. Core Flood Tank Tanks)
? (Cont.) Soron Concentration Verifica-I tion Core Flood Tank Power Source Check - Phase II l-High Makeup and Purification System Test operability of HPI ar.d M/U Perform the High Pressure SFAS Test. Makeup Pump Post- pumps Injection (HPI) System Pressure Functional Test Verify Injection Rebutid Baseline Test Test and Inspect System Valves Injection Flow Rates i
NPI/ Makeup Pump Venting Venting of ECCS Piping.
Venting of HPI Pumps HPI Pump fi performance Test HPI Valve I
' Operability Check ,
4 1 HPI Suction Reverse Flow test I HPI-TPO-1 ECCS Verification of Flow Path HPI System Pressure Isolaton 2 O Test i h N Remote Shutdown Monitoring j
Instrumentation Channel Calibra-tion j t '
Makeup Makeup and Purificatton System Makeup Flow Tests at High Pres- Makeup and Purification tetdown and Functional Test. Makeup and sure System E.S. Test. Operational Puriftea- Purtftcation System Operation Checkout
] tion Test Makeup and Purification System 3 1 j SFAS Test Post Hodification Letdown Makeup Valve Air Storage Tank i Letdown Filter Functional Test Filter Backflush Valves Test Leak Test C Intermediate Cooling System
! Operational Test to Verify Cooling Capaht11ty of Makeup /
' Letdown j (See HPI System for Pump .
4 Testing) l Containment AB Integrated Leak Rate and Containment Butiding Integrated ,
4 J Building Pressure Test Leak Rate Test. Satisfy l
' requirements of TS 4.4 (LLRT i
' underway; ILRT scheduled inspection & RB Isolation Valve) ;
j
, a
} r
- s. .
SYSTEH RANCHO SECO (ORIGINAL STARTUP) RANCHO SECO (RESTART) DAVIS BESSE (RESTART) THI 1 (RESTART)
Containment Containment Building Penetra- Containment Vessel Lucal Leak Butiding tion Leak Rate Test Rate (Cont.) Hydrogen Purge System 18 Honth Hydrogen Purge System Refueling Interval Test or Special Test. Hydrogen Dilution System Quarterly Test Auxiliary Turbine /Hotor Driven Auxiliary AFW Pump Functional Tests AFPT Main Steam Testing. AFPT Verify the two motor-driven Feedwater Feed Pump Functional Test. Adelssion valve Test on EFW Pumps and the Turbine Motor Driven Auxiliary Feed Auxiliary Steam. AFPT Admis- driven EFW Pump all start on Pump Functional Test ston valve and Governor Cold loss of all RC pumps Testing. Auxiliary Feedwater System Monthly Test. AFPT Supply Pressure and Valve Test.
AFW Pump Room Ventilation Test AFW Valve Functional Tests AFW Pressure and Valve Test AFW Flow Path Verification Test Auxiliary Feedwater System Refueling Test j3 AFW Flow Meters Calibration U$ AFW Actuation Signal Calibra- Auxiliary Feedwater Valve Logic tion Verification. AFW Steam Line Break Alann. Integrated SFRCS AFW. Auxiliary Feedwater Pump Train 1-1 Instrumentation '
Functional and Calibration Test.
Auxiliary Feedwater Pump 1-2 Instrumentation Functional and Calibration Test AFW System Channel Functional Test AFPT Overspeed Trip Test and AFPT Governor Low and High Speed Stop Verification Steam Trip Survey (AFPT Supply Line Steam Traps). AFW HS Supply Line vibration Test.
ADFP Acceptance Test. HOFP Hotor Run-IN. HDFP 18 Honth Test Auxiliary Feed System and OTSG Control Test
4 i
APPENDIX E
- SAMPLE PORTION OF l l
ACTION PLAN ACTIVITY TRACKING REPORT i I i
3 July 1986 1
6 4
'I
}
l l
4 l
I
..-_m..,__- - - . _ . . . . , , _ _ _ - , , _ _ - . . ,_ _ _ _ _ , .. _ _ . _ , . , _ , _ _ _ _ _ . - . . . . _ ,m~m_-._ _ _ _ . , - __ , . - . . _ ,_ _ . ..,.____ __ _ - _ _ _ _
I i
b' RECDP9tENDATION LIST (x
columin Title Descriotion Enlumn Title Descriotion i
l l
Log No. RRRB identifier defined as: Val /Inval defined as: '
1$.XXXX section 15. 12/26/85 Transient IR - Invalta-Redundant l Action List Recommendation 16.XXXX NRC Region V Recommended Action List 10 - 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 $taff Interviews Otsposition Organization assigned by RR8 to 22.XXXX Deterministic Failure Consequences Organization investigate and dispose of 23.XXXX BWOG $ top Trip Program proposed recomenoation:
24.XXXX RRR8 Observations CH - Chemistry (these items are not recommendations but EM - Electrical Maintenance but may lead to recomendations.) proposed recommeno.? ton:
25.XXXX Dept. Managers Reconmendations IC - I and C Maintenar.a LI - Licenstng NE - Nuclear Engineering Recomendation Brief description of proposed reconmendations NO - Nuclear Operations RC - Regulatory Compliance RP - Radiation Protection Initiator Person initiating reconmendation TR - Training TS - Technical Support 01sp Eng 01spositioning Engineer j $Y$ System 3 letter identifier of applicable Oue Date Date disposition is due to RRR8 plant system. See AP.3. Enclosure 6.4 from Disp. Eng.
(additional designation for this report
- incidates disposition (if include a MPS Multiple Plant Systems: MGT - required) completed.
I Management NSA - No Systen Appitable)
$ch Cat Defined as RRR8 reconmendation for:
$U - Startup Cat Category - Area affected by recommendation N$ - Non Startup May include several areas and are defined: Undeterminded Status Results of RR8 review defined ast, NA - NOT APPLICA8LE PAG - Reconmendation has been i
DC - PLANT N00!FICATIONS forwarded to Performance Analysis Group a
MP - MAINTENANCE LL - LE$$0N$ LEARNED REC - Recommendation has been TR - TRAINING returned for Clarification OP - CPERATIDMS AND PROCEDURES OSP - Recommendation forwarned to EP - EMERGENCY PREPAREONE$$ Dispositioning body 00 - QUALITY AND QUALITY
- A$$URANCE
" ME - MANAGEMENT EFFECTIVENES$ Acted on Date Date reconmendation reviewed by CM - COPetITNENT MANAGEMENT Comments RRRB CO - CONFIGURATION MANAGEMENT RRR8 Coments HP - HEALTH PHY$1C$ RO - RECORO$ AND DATA BASE P91 - MATERIAL MANAGEMENT MANAGEMENT i
i s E-1 i
i i
w --, -,. .- -y , , ,- wv. ----- -m -- - ~ - - - . - - - , --- - . - - - - - , - --- --e-.- - ---- - - - - - . - - - - - - - - ~ w g r n-- -
Ragello. 1 87/01/06 '
RAM IECGO B ERTIGl LIST 15800 !ECO ltI1 EAR IEEMTDEi STATIGI IX4IT GE
PIDlIDE APPMPRIATE SIZE PIETECTIVE ICS DC V DEIRIIEi ! PAS 86/12/86 15.0106 FtE FOR IDE ILTagglTIldi CINENT 18 REED LORD CIET R301 dei KIIEDI IlfTElilulTED CGITML SYSTDI CABilET H41CO2 M CMILE H1R!.
DEME PMPER CIERDDflTIGI BET)EEN TE FUSE N UPSTEftl BREAMER. ILSO DEiURE Pfl0PER C00RDINATIGl BET)EEN ENSTIIEi REES Ale lpSTREAM Pfl0TECTIGl. (ACTIGiITDI3.F.8 l CLOSUIE REPORT). l i
15.0245 WFY M PUIli TO IEPUE M ICS DC V DEIRDE 1 PAS 86/03/86 EIISTDEi 30A BUS S1J ICS FEDER NEiWER llITH A 40A BIEMER (DEiDEERIMi DDEE IETICE R-9469).
(GGl ITDI 3.B.3, CLUME
!!EPORT). ,
15.0247-1 PulF0181 IOK IEIUllED BY EDI ICS DC V DEIRIIEi 1 PIEi 06/13/06 R 4359. DEiDEERINE D5 BEE EE R-03S9 CHAIEES TE ELECTRICfL SENSIIEi SOUKES FOR 1)E PGER SUPPLY MTUIS IN TE ICS M 1911 CABDETS. (ACTIGtITDI3.F.6, CUEUE lEPORT).
i 18.8045 EVALUATE M UllE (F A SIIELE IGER ICS DC V DEIRIMi 1 Pali 06/16/86 SJPPLY IDl! TOR FOR ICS, AS TE SIIELI MGilTOR LElWES M SYSTDI E-2 i
Page No. 2 87/01/3 1555 IECGOGSATIGi LIST 15800 !E!Il IEnBut IBEMTIldi STATIGI 1911T GE
- ++ DEIBRER 26,1985 TIUBEIENT **e4+He#9
, LGi YAL/ DISPOSITION DATE IEL IEIDeseRTIGEi SYS CAT IIML OlgulIZATIGl PRIORITY STARS ACTED Gi Vl1JEINELE TO SIIELE FAILUE.
15.geW OPERATIGEi SELLD MFY ICS 19 Y OPEMTIGIS 1 PAS 86/16/E APPIEPRIATE PMCDulES BASED UPON ACTIGt ITEM 1.J fue 3.E.1, CLOSulE IEPORT. 001 E-127). TE E-DEllilIATIM STATLE & SPECIFIC j ICS EXIJIPIENT IS IEEDED TO WFY TE ICS RECOVERY PROEDURE.
15.8150 A CMil#LTY PRIEEDUIE SOLD E ICS OP V OPEMTIGEi 1 PAS 85/27/E DEVELOPED TO PlWYIE ElIDIUCE TO TE OPERATullS W LOSS & ICS IDER.
THIS PROCEDUE 90LD BE ilRITTDI TO ENSulE E.94 OR E.E5, IE APPLICMLE, HRVE IEEN CGSLETED TO TE POINT THAT llCS PRESEUIES 18e TBSERATulES AIE STABILIZED Ale lpSER DOO6TMTED OPEMTOR camel BEFulE ICS TR0lEILE900TIldi IS CGOENCED. TE PROCEDUIE FOR ICS POER RESTOMTION SDLD STIEIS CLOSINS S1/52 TOETER NOT SEPAMTELY. (NRC/ stb E ETIIEi 2/10/E). MFY LOS 815.8137 TO BE PART OF THIS IECGOENDATION.
(ACTIGl ITEM !$, (105UIE IEPORT, SECTIGtEll.
15.0063 LOSS & INIDil51TED CGlilEL SYSTDI ICS DC, V OPERATIONS 1 PAS 35/30/E PROEDURES SOLD DIECT TE OP OPEMTORS TO IUCE TE ATIEIiPERIC Dlfp VfLVE 130 T15tBIE BYPASS VALVE E-3
page No. 3 87/01/ 3 RRHB IEIDOEleATIGI LIST 15800 SECD ltIIIAR IEIERATIldi SMTIGl (MIT GE
- +e DECDeER 26,1985 TIUudilENT e+e++++e+e s
VIIL/ DISPOSITIGl MTE LIE SYS CRT INWE. OlBWIZATION PRIORITY STATIE ACTED (31 le. IECOOGSATIGEi IIJTM1IIEE SWITDES IN TE 'CLDSE' l
IEIE, fee TE AUIILIAllY FEED WATER
- L Ale N FLOW CGiTIEL CGITELEllS IN l
'futeAL' E91E ATTDIPTIldi10 IE-ESTABLISH POER TO TE l
INTDillATED CGITIEL SYSTOL ONCE IliTDIRATED CONTEL SYSTDI P0lER IS STABLE, TE VILVES CAN BE ETINDED TD M INTDIRATED CO W OL SYSTDI BY BOINil TO 'lODUL' Ale ' AUTO' RESPECTIVELY.
1 EPAIR TE INPIEIPDLY TBDillflTED ICS 19 Y OPDIATIGEi I PA6 N/02/E I 15.0287 WIE IN TE INTEBMTED CGITIEL J SYSTDI C1818ET. (CGIPLETD (ACTIGI ITBI 2A Il1XIT CIEEE IEPORT). I IESILVE M CALEE F TE DIFFUEIEE ICS 19 V OPDulTI(BEi I PA6 06/02/E !
15.0290 '
IN M M rGae" R UES F M l
TIIE DEUlYS (31 SI fue S2 (ICS P0lER SUPPLY IOIITOR SWITDES)
WITH M DESIliN VALLES. (ACTIGl ITDI 2A IWOT CIEEiE IEPORTI.
5.0001 IN(2.UDE ' LOSS F INTDiMTED (DITEL ICS TR V TilA!NIIEi I PIE 05/30/E i SYSTDl' IN TE TMINIIEi PIEMilWDL THIS TMINIMi WILL AID TE OPERATDRS DURIN6 A 'PARTIfL LIES F INTDiMTED CGITIEL SYSTDl' A6 lELL AS TUTR. LDB6 7 INTilEliMTED CG(TIEL SYSTEM CAEUALTIES. (ITDI IC, ACTIGl LISTI.
- l l
l E-4
Pa08 No- 4 07/st/n IIIS IEIDIEISATI(Bl LIST NIOC !ECO ltRIEAR IBEMTIIEi STATIGI INIT M i+ese**** DECD EER 26,1905 TRBEIDIT **+++e***e LOS R / DISPOSITION DATE
- 10. IECDOBERTIIDEi SYS CAT IIME. OE MIZATIGI PRIORITY STATUS ACTED (31 i
i 15.0098 EMURTE TE AUZILIAllY STEM ICS, DC V DERIDEi 1 PA6 06/02/E IEElldi RVE FETI(BI FDR LOSS F RSC INTEliRATED CDITIEL SYSTBI DC POS g usasvi INROILITY TO CGmEL AUIILIARY STEM PRESEUE M IlWLDENT CORIECTIVE ACTI(M Ni IEEESSARY. (ACTION 1791 1.J Ale
. 3.E.1 (2.MiUIE IEPORT).
15.0099.A PIENIE A B CDmEL STATION IN ICS, DC V DEIR!lgi 1 PA6 06/02/E TE CDmEL IEDI (31 H1R1 TO: ILlal NSS lEITGIITIC CLDIME F TE ATNOSPERIC Dlfip RVES/TURBIE BYPASS R VES.
15.0099.B Pil0 VIDE A B CDGEL STATI(Bl IN ICS, DC V DER!lgi 1 Pali 05/3B/E TE CDmEL IEDI Gl H1R1 TU: MSS PIENIDE ISIONL CLGil#E F TE ,
I IENS/TIMi.
15.0099.C PENIE A 161 CDmEL STATI!BI IN ICS, DC V DEIRIlgi 1 PR6 06/02/E TE CDmEL MDI WI H1R1 TO: TO LEI 6 (NDIRIE TE R11DOITIC CURiulE F TE ATIESPERIC Dlflp RVES/TUR0!E BYPASS VALVES AIS ALL(ni TE l
OPEMTOR TO OPEN TE ATIESPERIC Dl#1P RVES OR TullBIE BYPRSS VILVES (50 PDICENT OPEN) ELE INTER 61mTED C0itTR(L SYSTDI PGER IS LOST. (ACT!WI ITEN 3.F.4 Ale 5 (LOSURE ilEPORT, IEFDBCE ENSIIEERINii CHmE 10 TIE R-6357 A/B).
, 1 l
j 3.0009 BIE6 STOP-TRIP PMSNet TRIP ICS, le V IIIINT 1 //
i l E-5
page No. 5 57/01/3 MulB IE:(BUGeATI(BI LIST IUDOE !ECO IGELEAR IBEIRTIIEi STATIGI LMIT (BE
- DECD W I 26,1985 TINUEilENT **********
1Ri WIL/ DISPOSITI(Bi DATE le. IEcoleseATIGIS SYS CAT IIME. OIBulIZATIGI PRIORITY STATUS ACTED (31 l
EDIETItal IECDeGEATIGl 1911 l TIHeb!CS:
IlWIENDelTS IN ICS TISE CalTilIL Clit:UITS.
23.8011 BWS STOP-TRIP PIEXilugl TRIP ICS, IIP V MINT 1 // i IEETIM IECGOGeATIGI fel! ;
TR-til-ICS: DETDIIIE IF TE liRID l j
FIEEENCY ERIUll CIllCLIIT HAS BEDI DETLSED.
15.8145 E.82.14 SOLD E MIEIFIED TO ICS, OP V OPERATIGEi ! PAS 86/02/86 1
IEFDEG ICS PG0t AS IELL AS 1911 1911 IGER IF A CRE5LTY PMCEDLNE IS TO BE DEVE1DPED FOR LDSS F ICS PGER.
A Vit.IDATIM fee 'SIFICATIM IEVIEW 90LLD IE PDFulBED WI ATOS PROCEDUlut ElIDELIES leDIESSIIEi LDSS & ICS P0lER.
23.0008 BlES STOP-TRIP FROSilful TRIP ICS, OP V OPEMTIGdi 1 //
IEETIM IECDeetRTIGI TlH06-ICS 1911
- IIIPIENDENTS TO IEACT0lt IUeR(X UIPIBILITY.
218012 BlES STOP-TRIP PlelilWWI TRIP ICS, OP V GPERATIGEi ! //
IEETIM IECGGGEATIt31 IMI, TR-412-ICS: DETElIIIE IF OPDIATDR ANS ,
HAS IEEESSARY llF0lWIATIGI F1DI PIlOCEDURES, IIGICATORS, ETC., TO KTECT LDSS F *! PWER.
l i
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page No. 1 w/e! G IMS IEIDOEleATIGI LIST .
18800 !ECO IGREAR IBEMTIIEi STATIGI MT (BE
++******* DECD OER 26,1985 TIBUEIENT **********
i LIE VltJ DISPOSITIGI DATE
- 10. IEDeseATIGE SYS CRT IIME. OlselIZATICBI PRIORITY STATUS ACTED GI 19.0836.A CDELT A TIEUB4 IEVIEW F ICS, OP V OPEMTI'1B 2 PAS 05/30/86 AP.100-199, PIEEESS STIWSARDS, M EW IIENTIFY M CORIECT ERIERS, TO VERIFY EADI SETPOINT. GCE VERIFIED, TE SETPOINTS SIGLD E DEDED AEAINST ACTIR. ETPOINT IECORDS FOR TE DEVICE. TE IEVIEW SHulLD PROCEED ACCORDING TD A PRIORITIZED LIST F SECTIGIS.
TMEE SECTIONS INVILVIIE SEETY llELATED AIG VITAL EQUIPIENT SIGLD E DOE FIRST.
23.8004 FOR TE TURBIE EADER PIESEUIE ICS, DC V 2 //
INPUT TO TE ICS, A IGIFICAT!!BI 1911 SIGLD E INPLDENTED TO AUTDWITICALLY DETECT AN IBMLID INPUT fue SWITDi TO ITS IEIUSANT C M TERPART. TIE CGEPTtst.
DESIDE FOR THIS NIEIFICATIGi AIE DESCR! LED IN !ECTIGI 3.4 7 TE SMCE DOCGENT. GE DESIliN IS APPLICIELE TO PUDITS l#1VIIEi IDE i
THAN DE EADER PIESEUIE SIB 88. PER LOOP AIC TE OTER DESIIBl IS APPLICABLE TO PUuffS l#1VIIEi (BLY OBE SIBIAL PER LOOP.
23.0001 TE ilC FLOW INPUTS TO TE ICS ICS, DC V DEIRIls 2 //
9OLD E DELETED FA IEPUIED WITH 1811 EIRJIVlt.Dif SIDELS BASED ON IC PlplP STATLE IN ACCORDl8E WITH EITER F 1
TE TWO CulCEPTUAL DESIliNS PRESENTED IN TE SME DOCGENT.
CulGRRENT WITH TE IIEPUEBENT F RC FLOW SIENALS WITH EEVII.DIT SIIBIALS, TE EZISTDEi LLD LINIT l
E-7 I . .__ _ _ ._ - - . - _ _ . . _ . _ _ _ _ - -. __ _ . .- _ _- -. --
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07/01/86 M IEDGEleATIGl LIST '
15u00 !EID IRI2. ERR EElEllRTIIEi STRTIGl LMIT GE
++ Hue *e IEEIER 26,1905 T153EIENT **++e****e IAS YlR1 DISPOSITIGI DATE STS CAT IIML GuiRNIZATImi PRIORITY STATIS ACTED GI
- 10. IECGeseRTIGE
,,ED . RC Ft SmD mmD.
ICS, CD V EldiRIIS 2 PAS 35/27/86 23.0002 BW6 STOP-TRIP PIE 10 Aft 1 TRIP IEETIGi IEGOeeRTION THee-ICS lel!
FOR T-iET fue T-OLD, A ,
I IGIFICATIIDI 9OLD BE IlgUJENTED TD fRITGWITICILLY DETECT ful IIMLID INPUT fue 911T01 TO ITS REDLIDIli C(UffERPRRT. M CDEEPTINL DESIM FORTHISIEDIFICATIGiISDESCRISED IN SECTION 3.2 F M SWRE DOC 3ENT.
23.0003 506 STOP-TRIP PlWSIRI TRIP ICS, DC V DEIRIldi 2 //
IEETI(M IE30GeRTIM 1911 TH003-ICS
- IEMOYE STARIUP FW FLGi CDRIETIGi TO 15 TIN FW FLGi RfCTIGi FIDI M ICS.
23.0005 BW6 STOP-TRIP PIElil581 TRIP PlWBlugl ICS, DC V DeRIBEi 2 //
- IEETIM IEDeseATIGI 1911 T H I5-ICS. leEVE lelTIDI FLUI SIG5LS AUCTIDIEERIldi CIIE111TRY ,
FIDI RPS IIe IELOCATE IN M ICS.
23.0006 5i06 STOP-TRIP PIEEIRAM TRIP ICS, DC V DEIRIIE 2 //
IEETIGI IECDeGGRTIGI 1911 TR-806-ICS. DELETE FW TD5GUITINE CORIECTI(M TO FW DDRG FIDI ICS.
23.0007 506 STOP-TRIP PROGRful TRIP ICS, DC V DERIldi 2 //
IEETIM IEGeseATIGt 1811 T H 07-ICS. le UWE BTU LIMITS FIDI ICS.
I 23.0010 906 STOP-TRIP PIEMiluBI TRIP ICS, DC V DEIRIBEi 2 //
1 IEETIGI IEIDOGERTIGI 1911
- TH16-ICS
- ICS CDGIEL CIIEUIT IEEIFICATION.
CDEEPTIUL DESIR 7 lel Tave-CGITIELLED-BY-FEEDIRTER i
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Page No. 3 l 57/01/ 3 RRIB IEC(BOEleATIGl LIST l
I 15500 ECO IRELEAR IBElulTIMi STATIM LMIT SE t
. ++++++++e DEIB EER 26,1985 TIUUEIENT **++++e+++
LIE El DISPWitTIGl DATE IIME. ORBMIZATIGl PRIORITY STATUS ACTED ON W. IEIDeseATI(MS SYS CAT CIRC 1!!T:
Tne (BVIIIJS FEAT 13ES TO E '
IICORPORATED INTO M IMPROVED Taw EluCR COUR SDBE AIE M F(LLOWINS:
- 1. 'JSE (F A DERIVATIVE-PICPORTImut.-INTEBR CGmH1ER SO THAT FW F1.(NilATE CM E REDLEED AS Taw EMOR STARIS TO DECEASE lEYule ITS HISEST VILLE OR FW FLOWilATE CAN BE INCREASED AS Taw ERROR STARTS IICREASIIE FlDI ITS MINIIRM R1E. THIS WILL SWRTEN M 30 SEIGC LAli TIE M .
WILL llEDlEE M MPLITIEE M PERIDO F M OSCILLATI(31 IN Taw M llEACTOR POER.
- 2. INPUT M Taw ERil0R S11358.
DIRECTLY INTO M LOOP A M B FLOWRATE VERSUS DO W S COMPARATDRS O\ (Sl300lS) S0 THAT M ERlWR SIIBIL WILL CRilE GR.Y Tif MAIN CSITIRL RVES TO AIMUST TO REDLE M ERROR.
IEFER TO FIRJRE 7.31 FOR A CDEPTR AlliUMBENT FOR M IGi Taw-CDGil(LLElHlY-FEEDWATER CIRCUIT. M PID Ctem n 1ER C3 1D REPUG M SlsgER, INTEliRAL, M rCSITIVE ERIGI IEDLLES, M M J SIGNAL FTER T)E EIISTIIE RELAY (T) l IGLD BE IE-RQJTED TO T)E Flai ERROR COMPMATOR IGLLES IN LDOP A l 1
M B FW CGEIG. TMINS.
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page No. 1 87/01/06 RMB EIDeGSATIGl LIST N300 !ED ltI2. ERR fBGATING STATIM INIT DE e+e+++e+e DECD W t 26,1985 TIutSIENT ++e+++e+e+
tai Vlt) DISPOSITIGl DATE
- 10. IE De G CATIONS SYS CAT INWE. ORBANIZATION PRIGIITY STATIE ACTED (31 ESTAR.191 A PREVENTIVE MINTD63E ICS, NP V OPERATIGEi 3 PAS 86/p/86 15.0296 PIII:EIME 10 DED( TE OlGulTIGi F loll TE POER SUPPLY Ith!TORIIEi SYSTEM IN TE ICS fue 1911, IIE11BIldi TE TIE DELAY F SI Ale S2. (ACTIIM ITEM 2R 1100T CAUSE REPORT).
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j PPENDIX G RANCH 0 SECO R. IMPLEMENTATION ORGANIZATION i
SYSTEM STATUS REPORT ,
4 j NTROL ROOM / TECHNICAL SUPPORT CENTER ESSENTIAL AIR CONDITIONING SYSTEM July 3, 1986 i
t 4
SYSTEM ENGINEER RECOMMENDATION FOR RESTART J
i TEST REVIEW GROUP ACCEPTANCE OF TESTING
- 4 b
PAG ACCEPTANCE OF SY F R RESTART i
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SYSTEM FUNCTION The Control Room / Technical Suppor enter (CR/TSC) Essential Air Conditioning System is designed to provide a s le environment for equipment and station operator comfort and safety.
During certain abnormal ehts, as noted below, the CR/TSC Essential Air -
System is automatica11 tYated and started. The CR/TSC Essential Air Conditioning System prised of two 1007. capacity redundant trains. Each train is compri o e following:
b Esse al 1tration Unit consisting of a moisture eliminator, electric d ct er coil, two HEPA filter banks, two carbon filter banks, and a te fan.
Essential Air Handler consisting of a medium efficiency filter bank, direct expansion cooling coll, and circulation fan.
4 Essential Condensing Unit consisting of a reciprocating refrigerant compressor, air cooled condensing coils, condensing fans, receiver,and associated refrigerant piping and valves.
4 Fourteen dampers and associated ducts.
The CR/TSC Essential Air System performs the following functions:
- 1. Isolation of the CR/TSC from potentially radiologically contaminated air 1
lC/) during a radiological event by automatically closing sixteen normal air conditioning system isolation dampers (eight isolation dampers are actuated by each train) and opening twelve essential air handler isolation dampers (six are actuated by each train).
- 2. Isolation of the CR/TSC from air potentially containing to gas during a toxic gas event by automatically closing eighteen isolati ampers (nine isolation dampers are actuated by each train) and%ning ten essential air handler isolation dampers (five are at W by each train).
- 3. Provide cooling for the CR/TSC during radiolog- 1, toxic gas, CR/TSC high temperature, or loss of offsite powe n s by maintaining the CR/TSC temperature at at temperature 80 r les:.
- 4. Prevent inflitration of potentia ra ologically :ontaminated air into j the CR/TSC during radiological e by pressurtzing the CR/TSC to 0.125 IWG relative to outside at here.
- 5. Provide fresh, filtered, conditioned supply air for CR/TSC ventilation during radiological and loss of offsite power events by maintaining a
! flow of 3,200 cfm through the Essential Flitration Unit.
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. 4 REFERENCE DOCUMENTS REVIEHED
\ 1. Control Room / Technical Supp t Center Habitability Study i 2. Rancho Seco Technical S f Nations 3.13 and 4.10 l
- 3. Engineering Change ice A-3920
- 4. Engineering C otice A-3660
- 5. Ellis and hs CR/TSC Essential Air Conditioning System Operating and Main anc@ Manuals, M13.15-103
- 6. te Safety Analysis Report, Sections
- 8. Troubleshooting Action Plan 1.R.1.and 1.R.2
- 9. ANSI N 509
- 10. ANSI N 510
- 11. ERDA 76-21
- 12. American Air Filter Air Flitration Untts TSC and Control Room; l M13.16-IM01 i
l 13. US NRC Reg Guide 1.52
- 14. US NRC Nureg 0737
- 15. Special Test Procedures STP-189, STP-194, STP-198
- 16. Non-Conformance Reports %
- 17. Maintenance Information Management System
- 18. CR/TSC Essential Air Conditioning System ni g Manual
- 19. Plant Operating Procedure A.14 4
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; I e
i RESULTS TESTING HISTORY REVIEN t
- j. (Later)
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fm SYSTEM PROBLEM CR/SC Toilet Area Exhaust Fan Iso tion Damper A (HV-54727) is not positioning properly when actuated and/or dam osition displayed by IDADS is not proper.
SOURCE: STP-194 da T leshooting Action Plan a.R.1; RRRB Action Item 19 10.8 NUMERICAL IDE :
DESCRIPTIO D in rformance of STP-194 (CR/TSC Essential Air Condition Safety ur s Operation and Flow Detection Test), HV-54727 was indicated by I to be in both open and closed positions simultaneously when the system was operated in the ventilation (high temperature / radiation) mode. 1 RESOLUTION OF PROBLEM This problem is being investigated in accordance with Work Request Number g 114655 and 114656.
Investigation conducted in accordance with Work Request Numbers 114655 and 114656 has revealed the damper does not consistently stroke thru full level.
! The damper blade is not hindered by binding or other interference which would limit travel. The damper actuator is being investigated in accordance with '
Os Investigation of the discrepancy is to continue until a root cause is Identified and the damper functions properly and consistently.
Proper function of the damper and the associated IDADS damper position indication will be vertfled by performance of STP .
i Proper function of the damper will be periodically checked b e formance of SP.211.01A.
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SYSTEM PROBLEM O- Shift Supervisor Office Exhaust Fa Isolation Damper B (HV-54732) is not positioning properly when actuate /or damper pos! tion displayed by IDADS
,is not proper.
SOURCE: STP-194dakTroubleshootingActionPlanI.R.1; RRRB Action Item .0M O. B NUMERICAL ENTlWIER:
t>
DESC ION Okng *]erformance of STP-194 (CR/TSC Essential Air Conditioning Safety.
Feucres Operation and Flow Detection Test), HV-54732 was indicated by IDADS to be in both open and closed positions simultaneously when the !
l system was operated in the ventilation (high temperature / radiation) mode.
l RESOLUTION OF PROBLEM This problem was investigated in accordance with Work Request Numbers 114656 i and 114657. The damper was found to function properly. The damper position limit switches, which are the source for the IDADS damper position indication, i
were found to be operating intermittently. The limit switches were adjusted.
, IDADS indication of damper position was subsequently checked and found to be
, proper.
Proper post maintenance function of this damper and the IDADS damper position
- indication will be verified by performance of STP .
4 Proper function of this damper will be periodically checked by performance of Surveillance Procedure SP.211.01D. ,
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SYSTEM PROBLEM p) i V CR/TSC Essential Filtration Unit low Transmitter (FT-54701) and/or related IDADS components:
- 1. Operate intermittent a r provide intermittent flow indication on the IDADS monitor. -
- 2. Control flow a te which is in excess of the Technical Specification 4.10 flow mit Wnd/or indicate a flow rate which is in excess of the Technical ification 4.10 flow limit.
R c. ' December 26, 1985 plant transient data, STP-194 (CR/TSC ntual Air Conditioning Safety Features Operation and Flow Detection T ; Troubleshooting Action Plan I.R.1; RRRB Action Item 19.0010.8.
NUMERICAL IDENTIFIER:
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 booster fan inlet vanes in the unit and inputs to the IDADS computer. The computer then processes this signal into an equation which prints out the flow rate in cfm.
During performance of STP-194, IDADS indication of flow through the s 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 CFM; within Tech Spec 4.10 Limitations.
RESOLUTION OF PROBLEM This problem is being investigated in accordance with ST 8 /TSC Essential Air Filtration Unit Flow Test) and Work Requ s mbers 114658, 115707.
Function of Flow Transmitter FT54701 was invest ed in accordance with Work Request Number 114658 and I&C Periodic Main nan alibration Task No.
04767. A cause for intermittent IDADS fl 1 ation was not identified.
Investigation of intermittent IDADS in ti is continuing in accordance with Hork Request Number 115707. -
Flow through the CR/TSC Essential Filtration Unit a (SF-A-7A), based upon measurements made by performing STP-198, is 3,010 cfm consisting of 1,507 cfm outside air and 1,507 cfm return air. Flow through the filtration unit indicated by IDADS during the test was 3,540 cfm. IDADS indicated flow rate through SF-A-7A was 530 cfm greater than the actual flow rate.
The computer equation used to calculate the indicated ficw rate wil? be revised to provide a more accurate IDADS indication of actual flow rate.
v Continuous and accurate IDADS indication of air flow thru the CR/TSC Essential Air Train A will be verified by performance of STP .
G-7
f-SYSTEM PROBLEM
\
CR/TSC Essential Filtration Unit low Transmitter (FT-54702) and/or related IDADS components control flow at te which is in excess of the Technical
- Specification 4.10 flow limit d/ Indicate a flow rate which is in excess of the Technical Specifica n 0 flow limit.
. SOURCE: Decembe 6, 985 plant transient data, STP-194 (CR/TSC Essential Air ioning Safety Features Operation and Flow Detection
- Test); Tr les ting Action Plan I.R.1; RRRB Action Plan 19.0010.B.
NUME L IEENTIFIER:
RIFTION:
Flow control and indication of the filter unit is accomplished by a Foxboro' Transmitter which monitors the differential pressure across the carbon filter and converts this into an electrical signal, which controls the booster fan inlet vanes and inputs to the IDADS computer.
, The computer then processes this 21gnal into an equation which indicates j flow rate in cfm.
During performance of STP-194, IDADS Indication of flow through the Essential Filtration Unit B (SF-A-78) exceeded Technical Specification 4.10 maximum flow limit. Actual flow rate was determined to be 3050 CFM; within Tech Spec 4.10 limits.
RESOLUTION OF PROBLEM
) This problem is being investigated in accordance with STP-198 (CR/TSC .
l Essential Air Filtration Unit Flow Test) and Work Request Number 115709.
l Flow through the CR/TSC Essential Filtration Unit B (SF-A-78), b upon i measurements made by performing STP-198,.is 3,050 cfm consist g 328 cfm return air and 1,725 cfm outside air. Flow indicated by.IDA r g STP-198 4
was 3,793. IDADS Indicated flow rate through SF-A-7B is c greater than i actual flow rate.
Return air and outside air flow rates will be adjust to 1600 CFM 210%.
~
The computer equation used to calculate the ind ed flow rate will be
! revised to provide a more accurate IDADS in cat of actual flow rate.
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SYSTEM PROBLEM O Operation and control of the CR/TS Operators.
Essential Air System is not understood by SOURCE: Operator i V ; Troubleshooting Action Plan I.R.2.
NUMERICAL IDENTI R 4
DESCRIPTIO Admitional training is necessary to enhance Operator knowledge /TSC Essential Air Conditioning System operation.
Clari cati n of backup power sources, various system operating modes, syjite ilability in the various operating modes, and system rf ces is necessary. - .
, RESOLUTION OF PROBLEM Additional training will be provided to improve Operator knowledge of the CR/TSC Essential Air Conditioning System. Topics to be presented include:
- 1. 01fferences between CR/TSC Essential Air Train A and Train B in the Interim Number 2 configuration.
- 2. Clarification of backup power sources for the system.
- 3. Various system operating modes.
- 4. System availability in the various operating mode.
- 5. System interfaces.
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SYSTEM PROBLEM 4 N Operating Procedure A.14 does not clude adequate instructions for operation of the CR/TSC Essential Air Syste SOURCE: Operator I rv s; Troubleshooting Action Plan I.R, RRRB Action Item 19 10.D NUMERICAL IDE :
DESCRIPTIO Oper Procedure 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 Unloading Scheme is not discussed.
- 5. Instructions for placing system in STANOBY are not complete.
RESOLUTION OF PROBLEM Operating Procedure A.14 and related procedures will be revised.
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m SYSTEM PROBLEM 4
Control Room pressure is not main ined at or above 0.125 IWG relative to outside atmosphere by operation o ther CR/TSC Essential Air System train.
. SOURCE: Interviews t -189 Coordinator; RRRB Action Item 19.0010.8 NUMERICAL IDENTI ER:V DESCRIPTI :
b Duri per rmance of STP-189 (Control Room / Technical Support Center Press Test), CR/TSC pressure could not be maintained above 0.125 IWG ti e to outside atmosphere by operation of either CR/TSC Essential A . rain.
RESOLUTION OF PROBLEM The following checks of the CR/TSC Essential Air System dampers will be completed in accordance with Hork Request Numbers 114662, 114663, 111572, and 114655:
- 1. Visual check of damper blade position in actuated and unactuated states
- 2. IDADS indication of actuated and unactuated damper positions.
Upon completion, perform STP 189 to verify the Control Room /TSC is pressurized
\/ to at least 0.125 IWG relative to outside atmosphere by operation of each CR/TSC Essential Air System train.
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I SYSTEM PROBLEM !
'A d Control Room /TSC Essential HVAC S tem (High Noise in the Control Room)
SOURCE: Meeting minute of HYAC Upgrade Meeting held 12/04/85 (J. Naleway/ D. Abb t stribution, 12/09/85); RRRB Action Item 19.0010A; Interview BN Laboratories Inc. Letters, 11/27/85 and 12/09/85 NUMERICAL ENT ER:
DESC TIO at on of each train of essential HVAC equipment produces high sound.
I s in the Control Room and impairs communication with the Control Room and while using the telephone. When both trains of equipment operate simultaneously, as occurs during automatic actuation of the essential system, the sound levels increase and have a severe impact on communication.
RESOLUTION OF PROBLEM The operating procedure for the HVAC systems presently specifies that if both trains are initiated by automatic signal, one train should be secured.
An evaluation of the noise and vibration levels of the essential system is presently being performed to establish the cause(s) of the high levels. Based on the results of this evaluation, immediate and long term resolutions will be pursued.
For restart, the following actions are recommended:
- 1. Evaluation and replace, as required, air diffusers to redu low velocity through diffusers.
For long term, the following actions are recommended: 9
- 1. Add additional balancing dampers to provide b a r distribution control.
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SYSTEM PROBLEM V Control Room /TSC Essential HVAC S tem (Refrigeration Problems)
SOURCE: Meeting minute of HYAC Upgrade Meeting held 12/04/85 (J. Naleway/ D. Abbo t stribution, 12/09/85); RRRB Action Item 19.0010C and 19.001 -
NCR S-5028 NUMERICAL IDEN :
DESCRIPTIO The c is of the refrigeration portion of the HVAC system do not it reasonable flexibility of the system to respond to all variations i ling loads with respect to ambient conditions. In particular, during low ambient conditions, the system-fails 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 OF PROBLEM
. A modification was performed in September 1985 (Reference ECN R-0142) which opened a flow path for refrigerant under low load conditions by opening a solenoid valve which had erroneously been controlled to close at low load.
Subsequent surveillance testing verified that the system was working satisfactorily.
s O For restart, the following additional actions are recommended:
- 1. Revise the controls for the condenser fans to allow cycling of two of the three fans in response to ambient temperatures. This will provide improved responsiveness of the condensing capacity to cooll load with regard to ambient conditions. This change is being implem d by ECN R-0769.
- 2. Provide a system pumpdown capability to return ref r to the receiver when the system is shut down. This ch ge 11 prevent damage to the compressor and enable more reliable st o the system.
For long term improvements in the operation of sy tem, the following actions are recommended:
- 1. The control of the loading and u d1Fg of compressor cylinders should be from suction pressure rather the temperature in the Control Room. This would allow a direct c trol of the compressor based on actual system loads rather than expected loads. ;
- 2. Modification of the evaporator control sequence based on actual
. operating experience to provide closer incremental control is desirable. This will require additional data to be taken during surveillances to monitor operating characteristics of the system.
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G-13 w -y pp - - - - + - w ---
SYSTEM PRJBLEM (Continued)
, O. Control Room /TSC Essential HVAC S tem (Refrigeration Problems)
~
- 3. The hot gas bypass' shou be Yontrolled through the evaporator to prevent gas through 1 Ive coil in conformance to good refrigeration pract .
- 4. A new P&ID sho developed to document the refrigeration system of
- the Contr Roo SC Essential HVAC System. This drawing will be in 4
greater de than the schematics provided by the vendor, and will faci e aintenance and troubleshooting of the system.
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G-14
SYSTEM PROBLEM
\j Control Room /TSC Essential HVAC ( pntrols and operating status indication in the Control Room.)
SOURCE: Meeting min ehHVACUpgradeMeetingheld 12/04/85 (J. Naleway/ D. Abb to distribution, 12/09/85); RRRB Action Item 19.0010E and 19. 11 NUMERICAL EN ER:
b DESCR7 TI0f ma al control and indication switch for the Control Room /TSC .
E tial HVAC System only provides limited information regarding the operation of the essential equipment. It does not have the capability to distinguish between SFAS, high radiation, or Control Rocm high temperature starts since all three are indicated by a single light. No indication is provided for the status of the compressor. Therefore, to verify operating status of the essential equipment, the operator must either check the remote local indication or review the output of IDADS.
RESOLUTION OF PROBLEM For restart, the following actions are recommended:
fw 1. Revise the operating procedure to address how to distinguish between the actuating signals prior to isolating both units. Additional operator
('v) training should be provided to promote a good understanding of the characteristics of the system.
For long term enhancement of information provided to the operators, the following actions are recommended:
- l. The control switch for the Essential HVAC System shoul e modified to i provide an indication of the signal which caused t ys to actuate.
- 2. An indication for the operating status of the essors should be added in the Control Room, either near the con switch for the system. ;
or in the IDADS computer.
- 3. EvaluatetheEssentialHVACSystemtkterminewhatothersystem )
operation indications should be e akT11able in the Control Room.
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G-15 l
p SYSTEM PROBLEM Control Room /TSC Essential HVAC S em (Damper and Damper Actuation)
SOURCE: Interviews; NC hS- ; STP-194; Troubleshooting Action Plan I.R.1; \Y RRRB Action Item 19 10.B.
NUMERICAL IDEN :
DESCRIPTIO C tr m isolation dampers have experienced difficulty in reaching
-c ose position and may be leaking. Access doors were not installed
^
n all dampers making maintenance of these dampers very difficult.
1 RESOLUTION OF PROBLEM Recommended actions prior to restart:
- 1. Initiate and implement a design change to add access doors where necessary for inspecting and maintaining the dampers.
7
- 2. Investigation of correct damper positioning and indication in actuated and unactuated states is presently being performed under Work Request Numbers 114662 and 114663.
{ 3. Evaluate the dampers to verify that they are installed in accordance with the manufacturer's instructions. Contact supplier to verify that the installation is proper.
- 4. Review and update preventive maintenance requirements for the dampers and their components.
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G-16
SYSTEM PROBLEM
) l
/ Control Room /TSC Essential HVAC S tems (High Vibrations on Condensing Units) l SOURCE: Interviews; Me ing YtInutes of HVAC Upgrade Meeting held 12/04/85, (J. Nalewa D. ott to distribution, 12/09/85); RRRB Action Plan 19.0010.L NUMERICAL IDE :
DESCRIPTIO O era of either of the Essential HVAC Systems causes a substantial at'on in the related condensing unit. This vibration has caused p straps to work loose and fall off and could therefore impact the integrity of refrigerant system if a pipe were to break.
RESOLUTION OF PROBLEM Maintenance did evaluate the noise level of the CR/TSC Condensing Units in December 1985 and found that one of the compressors was missing a 9 art which could affect the alignment. The compressor and motor for both units were realigned at that time and it did improve the condition.
Recommended actions for short term:
- 1. Perform additional investigation on the units to determine source of vibration.
O. 2.
Perform preventive maintenance on the units to routinely check the condition of all supports internal to the equipment housing.
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O G-17 l
,m SYSTEM PROBLEM l I d Control Room /TSC Essential HVAC S tem (Diesel Exhaust into HVAC Intake)
SOURCE: Meeting minute gof HVAC Upgrade Meeting held 12/04/85 (J. Naleway/ D. Abb t W istribution, 12/09/85); RRRB Action' Item 19.0010G; ODR No. 8 4 NUMERICAL IDE :
DESCRIPTIO Th e st of the existing diesel generators can be dispersed over the 11 ry Building area, including where the Control Room /TSC Essential.
H Unit outside air intakes are located. Concurrent operation of the diesel generators and the HVAC System could result in deterioration of 1 the carbon absorbent material in the filter banks due to the exhaust fumes being drawn into the unit.
RESOLUTION OF PROBLEM An evaluation of the HVAC OSA intake locations relative to the diesel exhaust stacks was performed and it was determined that there would be sufficient diffusion of the products of combustion in the exhaust stream with the surrounding air to avoid the problem of concentrated exhaust being drawn into the air intakes. In addition, a carbon sample was tested and the reduction in efficiency was not significant. Therefore, it was concluded that the diesel exhaust has a small impact on the efficiency of the carbon filters.
The following action is recommended for restart:
- 1. Perform an evaluation to determine what possible long term effects the diesel exhaust may have on the carbon filters. The refuel interval and 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> elapsed operating time procedures will continu sample the carbon such that any degradation will be detected. g b
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G-18
l SYSTEM PROBLEM O Control Room Normal'HVAC System ( Flow Balance)
SOURCE: Interviews, Ca at on Z-HVS-M0282, RRRB Action Item 19.0010.B NUMERICAL IDENTIFIE DESCRIPTION:
TheairbaheoftheControlRoomNormalHVACSystemhasnotbeen upda or verified since the original balancing was performed in 197, d pi veral modifications which could affect the balance or the heat
. s a result, the HVAC equipment may not be satisfying the current I ng in the Control Room.
RESOLUTION OF PROBLEM Long term recommended actions:
- 1. Initiate complete air flow balance of the Control Room HVAC System, including vibration testing of all rotating equipment.
- 2. Adjust and/or replace system components as required to support air balance (i.e., sheaves, belts, bearings, and thermal overload heaters).
This work has already begun.
- 3. Conduct a system review of the system design after the rebalance of the system to verify that system performance optimizes Control Room environmental conditions considering all the presently identified loads.
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-s SYSTEM PROBLEM Control Room Normal HVAC (Flow Sw ch)
SOURCE: Interviews;.M ing inutes.of HVAC Upgrade Meeting held 12/04/85 (J. Nalewa . ott to distribution, 12/09/85); SFR 280-85.
NUMERICAL IDENTI ER V .
DESCRIPTI -
1 D A pa -t' e low flow switch was installed in the cold air supply duct t s ff the normal air handler upon actuation of the Essential HVAC due to closure of the isolation dampers. During the colder .
p ds of the year, there is a smaller demand for cooling in the Control Room, such that the air. flow through the cold duct is decreased. This air reduction was low enough to cause the flow switch to trip and shutdown the normal air handler and lose all 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 OF PROBLEM A design change has been initiated to relocate the low flow. switch from the cold air duct to the return air duct where a steady air flow exists, independent of the heating / cooling load in the Control Room. This change is being implemented under ECN R-0314.
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G-20
l SYSTEM PROBLEM i
tem ControlRoom/TSCEssentialHVACSte(ElapsedTimeMeters)
SOURCE: Interviews, Te c Specification 4.10; RRR8 19.00118 NUMERICAL IDENTIFIE DESCRIPTION:
Technical h ification 4.10 requires sampling of the carbon filters in the entNil filtration after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of unit operation. Presently, no a tic monitoring device has been supplied to monitor the total
- hse time; therefore, operating time is estimated based upon Operator Lo # in the Control Room. Since these estimates are conservative, the carbon filters may be sampled more other than is necessary.
RESOLUTION OF PROBLEM
- A design change has been initiated to add elapsed time meters which will be
- connected to the breakers for the filtration units to monitor total operating i- time of each unit. This change is being implemented under ECN A-5782. The ,
surveillance procedures will be revised to check these timers to verify elapsed operating time since previous carbon replacement.
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i SYSTEM PROBLEM ControlRoom/TSCEssentialHVAC(%essurization SOURCE: Interviews; Te c eSpecification 4.10, STP-189 NUMERICAL IDENTIFI l DESCRIPTION:
Technicalhification4.10requiresthattheControlRoom/TSC
, Esse 1 m/AC System be capable to pressurize the Control Room and TSC 0 ING (min.); however, presently, no pressure tap exists to r the pressure in the TSC. Pressure readings are presently being. .
t by putting one tube under the door into the TSC.
RESOLUTION OF PROBLEM Recommended actions for short term:
- 1. Install permanent pressure taps through the TSC wall such that future pressure mecsurements may be made from a constant measuring point. It is recommended that one tap be placed through the north wall into the Auxillary Building main corridor and that a second tap be placed through the south wall into the NSEB Bridge so that measurements may be taken against both outside reference points.
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G-22
q SYSTEM PROBLEM Test agent injection manifolds n ssary for in-place filter leak testing have not been installed and tested.
SOURCE: NCR S4761 NUMERICAL IDENTI E DESCRIPTI -
d HEPA c bon filter banks in the Essential Filtration Units must be t riodically for leaks to satisfy Technical Specification 4.10 ;
ir ments. This testing is accomplished by injecting a test agent i 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 510, Testing of Nuclear Air-Cleaning Systems. Test manifolds have not been installed in the Essential Air Filtration Units.
RESOLUTION OF PROBLEM' ,
Install test agent injecticn manifolds.
l Test installed test agent injection manifolds to demonstrate test agents are uniformly dispersed in the airstream entering the filter.
O 4 $>+%
O G-23
SYSTEM PROBLEM Temperature in the Essential Unit pressor compartments may become excessive during hot weather operation.
SOURCE: RRRB Actio te .0010.8 NUMERICAL IDENTI DESCRIPTI W
t>
Vent to in Essential Unit compressor compartments is natural c cu on. During hot weather operation, temperature in the ar' ents are elevated when the system is in operation. The elevated t ratures may result in failure of the motor, compressor, or other component.
RESOLUTION OF PROBLEM Temperatures in the Essential Unit compressor compartments will be monitored during hot weather operation. Data obtained will be evaluated to determine the necessity for installation of a compartment air ventilation fan.
A compartment air ventilation fan will be installed if determined to be necessary.
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- G-24 L . _ _ .
SYSTEM PROBLEM
' Operation of the CR/TSC Essential ir Refrigerant Systems is compromised by certain piping and control proble SOURCE:
NUMERICAL IDENTI E !
DESCRIPTI -
b
- 1. l ea' in the CR/TSC Essential Air Train A refrigerant piping is ected.
~
2 A pressure regulator in the CR/TSC Essential Air Train A refrigerant system is subject to short cycling resulting in continuous surging.
- 3. Evaporator coil superheat should be checked and adjusted.
RESOLUTION OF PROBLEM Troubleshooting and maintenance as necessary will be performed to resolve the aforementioned discrepancies.
A Maintenance Procedure will be initiated to describe the steps necessary to w accomplish replacement of various refrigerant system components and to charge
] the system with refrigerant.
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lO G-25 i
,- SYSTEM PROBLEM
( The CR/TSC Essential Air Conditio g System is. subject to spurious actuations caused by high temperature in cer areas of the Control Room.
SOURCE: Occurrence sc lon Report; Septamber 7, 1985 NUMERICAL IDENTI ER.
DESCRIPTI -
d Temp tur sensors which actuate the CR/TSC Essential Air System in the e nt high Control Room temperature (80*F) are located approximately t eet above the floor in the vicinity of computer equipment.
T rature in this area is somewhat elevated relative to the average temperature in the Control Room. Spurious system actuation is caused by high temperatures in the vicinity of the sensors rather than excessive average temperature in the Control Room.
RESOLUTION OF PROBLEM A change in the actuation setpoint temperature should be considered.
Determination of the floor to ceiling temperature gradient.should be accomplished to enable establishment of an appropriate setpoint.
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_ . - . _ - ._~ .
-_. . J
3 SYSTEM PROBLEM
- (j) Air flow through the Essential Air Filtration Units may exceed the maximum rates specified in Technical Spec ation 4.10 for several minutes after the systems are actuated.
SOURCE: STP-198dahTroubleshootingActionPlanI.R.1; RRRB Action Ite 9.0910.8 NUMERICAL DEN ER:
t>
DESC TIO te fan inlet vanes, which control flow through the Essential .
F ation Units, are positioned in response to differential pressure j developed across the carbon filter banks. When the units are actuated, there is not 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.
j .
- l. RESOLUTION OF PROBLEM The Essential Filtration Unit Flow controls will be adjusted to reduce the time necessary to establish stabilized flow.
i Upon completion of adjustments, time required to establish stabilized flow will be determined for each train.
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G-27
, s SYSTEM PROBLEM N Surveillance Procedures SP.211.01 nd SP.211.01D do not provide a check of all possible damper limit switch ion indications. As a result, damper position and/or position indi o roblems may be masked.-
SOURCE: Troubleshoo g Action Plan I.R.1: STP-194.'
NUMERICAL IDEN :
DESCRIPTIO T re four possible damper position IDADS indications for each SC ssential Air System damper. However, only two of the four '
p ions are recognized on the pertinent surveillance enclosures. As a consequence, conflicting damper position indications are not detected.
4 RESOLUTION OF PROBLEM Survelliance Procedures SP.211.01A and SP.211.010 will be revised to recognize the four possible position indications.
A new P.M. will be initiated to verify proper damper function once each refueling interval.
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G-28
- _ - - - -,~w.-.---,.,,. , , . - - . - - - , - - - --.-%,---,,_,,m,r...-m----,..,r..-,, r m %,.
t SYSTEM PROBLEM Carbon filter tray mounting in th R/TSC Essential Air Filtration Unit must be evaluated for appropriate filt unting capability.
SOURCE: RRRB Action t .0011.D NUMERICAL IDENTI ER:V DESCRIPTI t>
The bon ilter tray mounting frames may be improved to facilitate !
main ce activities. j
- RESOLUT OF PROBLEM j 4
- A study of the carbon filter tray mounting frames will be performed to
- identify and evaluate improvements.
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G-29 ,
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SYSTEM PROBLEM ,
' l There is no approved procedure for alibration of the CR/TSC Essential Filtration Unit Flow Transmitter.
SOURCE: RRRB Action t .0010.8 NUMERICAL IDENTI ER:V DESCRIPTI :
b Diff ti pressure across the carbon filter banks is the parameter ms to control flow thru the Essential Filtration Units. The flow ac eristics are subject to change when the carbon is changed, I n sitating recalibration of the flow transmitter to assure the proper .
- flow rate is maintained.
! RESOLUTION OF PROBLEM A procedure for calibration of Essential Filtration Unit flow transmitters will be instituted.
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O G-30 i
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SYSTEM PROBLEM Work on fire dampers between the trol Room and TSC (currently deferred under MOD 111) is not complete.
SOURCE: RRRB Action m .00ll.A NUMERICAL IDENTI ER DESCRIPTI -
! t>
l Eigh is - ng dampers are to be modified and two new dampers are to.be inst
. Long term operation with compensatory measures _is not ra le. The ability to complete this work prior to startup must be e ated.
RESOLUTION OF PROBLEM The ability to complete the work before startup will be evaluated. If there is a reasonable possibility that the work can to completed in a timely manner, the work will be undertaken.
If the work is completed, surveillance procedure revisions and testing will be necessary.
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SYSTEM PROBLEM Quality of the existing CR/TSC Es tial Air System refrigerant compressor motors is questionable.
SOURCE: RRR8 19.0011 NUMERICAL IDENTI ER:V i
DESCRIPTI t>
The pre or moto'rs are ODP motors. Concern has been expressed about ty of the bearings, shaft keyway, and motor windings.
t q
- E ration should be given to upgrading these motors to TEFC or Mill a hemical motors.
- RESOLUTION OF PROBLEM An evaluation of the specifications for these motor
- will be conducted to l' determine the suitability of the existing motors for the intended service and the advisability of upgrading replacement motors for more severe service.
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G-32
n SYSTEM PROBLEM b)
Automatic initiation of CR/TSC Es tial Air System operation increases Control Room Operator work load d g certain critical plant operating phases.
SOURCE: on Plan I.R.
Troubleshoo$g NUMERICAL IDENTI ER.
l DESCRIPTI t> l Both ain of the CR/TSC Essential Air Conditioning System are a om ally started upon Safety Features initiation. The Control Room at r must direct attention to this system, decide which train to s own, then take action. This is believed to be an unnecessary contribution to Control Room Operator workload during critical phases of plant operation.
RESOLUTION OF PROBLEM A study will be performed to determine the necessity to start both trains of the CR/TSC Essential Air Conditioning System upon Safety Features' initiation.
The following will be considered.
- 1. Automatic start of only one CR/TSC Essential Air train in the event of a Safety Features initiation. Automatic start capability of the second train would be removed.
(. 2. Staging of trains to automatically start only one CR/TSC Essential Air tral- upon Safety Features initiation. The second train would automa*t ally start only if the first train failed to start or operate.
- 3. Removal of the CR/TSC Essential Air System automatic s capability from Safety Features. Neither train wo be automatically started by Safety Features actu n.
The nature of modifications necessary to facilitate f the aforementioned capabilities will also be considered. Modifications Y ich add undue complexity to the controls will not be acceptab 4+
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SYSTEM PROBLEM O\ The affect of CR/TSC Essential F1 CR/TSC radiation exposure levels ration Unit initial air flow rates on ot known.
)
SOURCE: USNRC conce l
NUMERICAL IDENTI ER DESCRIPTI b
Upon ste, actuation, air flow rates thru the CR/TSC Essential Air Fi tr n Units exceed the stabilized flow rates for several minutes.
g this period, the concentration levels of airborne radioactive p culate and or gas is assumed to be at maximum levels. The combination of the flow rates and the concentration levels may result in increased CR/TSC occupant exposure levels.
RESOLUTION OF PROBLEM
, A study to evaluate the affect of CR/TSC Essential Filtration Unit hi flow 1
rates and ht radiation concentration levels on CR/TSC occupant exposure levels will be completed.
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! G-34
- TESTING DONE
{U POST MODIFICATION:
4
- l. CR/TSC Essential A o Level Reduction ASpecialTesthocedurewillbenecessarytoverifythatCR/TSC noise level co Morm to the recommendations discussed in NUREG 0770 and that g capacity of the CR/TSC Essential Air System is not impa d he modifications.
d
- 2. e er Fan Cycling Modification A special test procedure will be necessary to properly start-up the .
modification and to verify that the CR/TSC Essential Air System performs properly during cold weather operation.
- 3. Paddle Switch Modification A special test procedure will be necessary to properly start-up this modification.
- 4. Test Agent Injection manifold Modification 1
A special test procedure will be necessary to verify that in-place filter leak test agents are properly distributed by the test manifold.
~ POST MAINTENANCE
- 1. CR/TSC Essential Filtration Unit IDADS Flow Indication i
A special test procedure will be necessary to verify t the discrepancy which caused an intermittent IDADS Indicat of flow thru the A CR/TSC Essential Air Filtration Unit ha en corrected.
- 2. CR/TSC Pressurization A special test procedure will be necessar verify the CR/TSC is pressurized to at least 0.125 IWG by ra ton of each CR/TSC Es,sential Air train.
4 4 G-35
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' TESTING --POST MAINTENANCE (Continued) l s_/ 3. CR/TSC Normal Air.Cond lo11ng System A special test pro dure will be necessary to verify the CR/TSC Normal Air Sys i rforming as intended.
- 4. CR/TSC Esse la iltration Unit Flow Stabilization A sp 1 Vst procedure will be necessary to determine the elapsed time een actuation and flow stabilization for each Essential ain GPERAB Y ERIFICATION Operability of CR/TSC Essential Air System train A will be verified by performing Surveillance Procedures SP.211.01A and SP.211.018.
Operability of CR/TSC Essential Air System train 8 will be verified by performing Surveillance Procedures SP.211.010 and SP.211.01E.
FUNCTION ABILITY VERIFICATION
- 1. CR/TSC Essential Air Conditioning System Dampers A special test procedure will be necessary to verify proper function of all CR/TSC Essential Air System dampers.
x_. 2. CR/TSC Essential Air Compressor Compartment ventilation A special test procedure will be necessary to obtain data enabling a determination of requirements for a compressor compartment ventilation fan.
- 3. Hi Temperature Actuation of the CR/TSC Essential A stem A special test procedure will be necessary luate the temperature gradient and distribution in R/iSC to establish a set point temperature for hi temperature tion of the Essential Air system.
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l TESTING (Continued)
ONGOING PERFORMANCE AND RELIABILI MONITORING Performance and reliability o th /TSC Essential Air trains A and B, including filters, fans, pe and refrigeration components, will be verified at least monthly performance of SP.211.01A and SP.211.010, respectively.
Performance and lia 1 ity of the CR/TSC Essential Air System dampers will be verified each r ing interval. A Periodic Maintenance Procedure will be necessary act plish this.
Perfor e nd reliability of the CR/TSC Essential Filtration Units will be -
verifle ch refueling interval by performance of SP.211.018 and SP.211.01E.
Performance and reliability of the carbon filter banks will be verified after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of elapsed operating time by performance of SP.211.01C and SP.211.01F.
Performance and reliability of the CR/TSC Essential Air System will be
- monitored quarterly in the Plant Performance Monitoring Program.
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4 l~ RESULTS OF REVIEW OF TEST RESULTS FROM DAVIS-BESSE (Later)
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, RESULTS OF REVIEW OF RECENT MAINTENANCE ACTIVITY ;
AND. MAINTENANCE HISTORY TRENDS (Later) I L
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_ _ _ . - __._____.-__-.._....____.~.___._____;._:.
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!. RESULTS OF SYSTEM MATERIAL CONDITION HALKDOWN
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APPENDIX H SAMPLE TEST SPECIFICATIONS i
The attached Sample Test. Specifications are representative of the test specifications which will be used in the test program.
Independent Control of AFW Control Valves Page H-1
- O i-Independent Control of ADV's and TBV's Page H-3 Fire Protection Valve Drain Line Page H-5 Functional Test l
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TEST SPECIFICATION O Independent Control of AFH Control Valves Revision 0 4
MODIFICATION NUC ENG ZANT ENGINEER NUC DESIGNATED ENGINEER 1
PURPOS -
To verify proper operation of the following:
- 1. Dual steam generator level lndicator on H2PS.
1
- 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-Oll.
Calculation for preset AFH valve bias for loss of ICS Power -
Z-FWS-IO105.
j PREREQUISITES: 1
- 1. Construction complete and turned over to Startup with ny ptions duly noted and recorded.
- 2. Ensure simulated loss of ICS does not adversel e et the plant.
- 3. Auxiliary Feedwater lined up to permit strok of the AFH control valves.
TEST METHOD:
I&C Maintenance Department to veri e following: !
- l. Installation 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.
- 4. Installation and operation of Steam Generator A and B Startup Level O Indicators LI-20503B and LI-205048 respectively. ;
The above will be verified by the following methods:
- 1. I&C Maintenance Departmen to verify installation and operation of instrument loops by calib g instrument loops using loop form per General Calibration P du e I-Oll.
- 2. WithICSpoweravkable,andManualControlStationsinautomatic,
, verify that So no1Y Valves FY-20527C and FY-20528C are de-energized and verify t lowing:
- a. Red cator lights above L&N Manual Controllers are not lit.
- b. Control Valves receive pneumatic control signals from ICS via
/P's by using existing Bailey Hand / Auto stations on Panel H1CO.
- 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 i
Auxiliary Feedwater Control Valve position for loss of ICS DC power. The valve position will be determined from the calculated AFW flow needed. Record the valve position at both AFW Valves L&N Manual Controllers (HC-20527 and HC-20528) indicator.
- b. Place L&N Controllers in manual and calibrate such that 0 to 100%
) Indication on the Manual Control meter corresponds to 0 to 100%
) Auxiliary Feedwater Control Valve stem travel. Verify "bumpless" transfer from Auto to Manual for the L&N Manual Controls.
- 4. Place the Manual Controllers in "AUT0" and remove simulated Loss of ICS Power and verify FY-20527C and FY-20528C deenergize a the red indicator lights go off. >
- a. Place L&N. Controllers in " MANUAL" and verify u ontrol as in t
3.b. Insure red indicator lights are'11t.
- b. Place L&N Controllers in "AUT0" and vert turn to normal ICS control by modulating AFW valves us thW Bailey Hand / Auto Stations on HlRC.
- 5. '
With Manual Controllers in "AU no simulated Loss of ICS Power verify normal ICS Control as i for each of the following conditions.
- a. Removal of AC power to the 24 VOC power supply for HC-20527 and j 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
- - _ ~ ._. _
t TEST SPECIFICATION Independent Control of ADV's and TBV's Revision 0
- MODIFICATION .
NUC ENG ZANT ENGINEER NUC DESIGNATED ENGINEER RT-UP COORDINATOR 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 H1RI panel.
j 2. Aux 111ary relay contacts in the ICS.
REFERENCES:
- 1. 'ECN R-03578 and associated drawings. ,
PREREQUISITES:
- 1. Construction complete and turned over to Startup with any exceptions duly noted and recorded.
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- 2. Ensure simulated loss of ICS does not adversely effect th ant. ,
! 3. Plant in cold shutdown and secondary side of Steam Ge tors
. depressurized or in a configuration to permit te g.
TEST METHOD:
, I&C Maintenance Department to verify the fo in :
- l. Installation and operation of the h nd TBV Override Switches, HS-20562C and HS-205618 respec elyUTocated in H1RI.
I 2. Hiring to ICS relay 86/ICS-PSM contacts.
The above will be verified by the following methods:
ADVs
- 1. With both hand switches HS-20562A (located on Shutdown Panel H2SD) and i
HS-205628 (located on Atmospheric Oump Valves Manual Control Panel) in
- " Auto" and HS-20562C (located on HIRI);in " Normal" verify
'H-3.
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- a. Solenoid Valves PY-205620, PY-20562E, PY-20562F, PY-205710, PY-20571E and PY-20571F are energized.
- 1. Simulate a loss o power and verify that the solenoid valves in 1.1 deenergize. Verify ADV's are closed.
' 2. Place the Override Switch HS-20562C in " Disable" and verify at he solenoid valves in 1.a. above reenergize.
i
- 3. ace \the ADV Override Switch HS-20562C in " Normal" and remove simulated Loss of ICS Power Condition, verify that the s lenoid valves in 1.a are energized.
4
. Place the ADV Override Switch HS-20562C in "Close", verify .
- that solenoid valves in 1.a. are deenergized. Verify ADV's 4-are closed.
i TBVs
! 2. W1th TBV hand swltch HS-20561A (located on H2SD) " Auto" and HS-20561B
- (located on H1RI) in " Normal" verify
- a. Solenoid valves PY-20561 A,- 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 s' witch HS-205618 in " Disable" and j verify that the solenoid valves of 2.a. above reenergize.
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- 3. Place the TBV Override Switch HS-205618 in " Normal" and remove i
the simulated Loss of ICS Power Condition, verify that the solenoid valves in 2.a are energized.
. 4. PlacetheTBVOverrideSwitchHS-20561 Bin"CQq,", verify that solenoid valves in 2.a. are deenergi . Wrify TBV's are closed.
i ACCEPTANCE CRITERIA:
- 1. Equipment tested and results meet cr ter nder Test Methods.
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i H-4
TEST SPECIFICATION Fire Protection Valve Drain Line v .F tional Test Revision 0 MODIFICAT NUC NG NIZANT ENGINEER t>
NUC PS DESIGNATED ENGINEER TART-UP COORDINATOR PURPOSE:
This modification includes the addition of a drain line from the TSC fire protection valve, through the switchgear room and corridor, to a floor drain located in the ventilation equipment room. The drain line will have a seal trap which, when filled with water, will inalntain the control room pressure boundary to preserve control room habitability as prescribed by Nureg 0737. This test will verify that the CR/TSC essential HVAC can maintain pres:urization of the areas in accordance with tech spec limits following the completion of this modification.
REFERENCES:
Ov 1. Tech Spec 4.10 " Control Room / Technical Support Center Emergency Filtration Unit"
- 2. NUREG 0737 Item III.D.3.4 - Control Room Habitability Requirements PRE-REQUISITES:
- 1. Grouting of penetration for drain line is comle
- 2. Water level in the seal trap is a minimum of s inches.
- 3. All other breached penetrations are fit wi h appropriate temporary-or permanent closures.
TEST METHOD:
- 1. Independently operate the "A" an "B" CR/TSC Essential HVAC Systems in the radiological /h1 temp mode.
- 2. Measure the pressure differential between the Control Room and outside atmosphere using the 1/2" dia. stainless steel tube.
- 3. Measure the pressure differential between the technical support center and the corridor north of the TSC.
A U
H-5
i j ACCEPTANCE CRITERIA:
i 1. The pressure of the Contr ~ Room relative to the outside atmosphere
- shall equal or exceed 1/8 g.
l- 2. The pressure of t Te cal Support Center relative to the corridor.
shall equal or ex d 1/8" w.g.
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