ML20198K899
| ML20198K899 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 06/02/1986 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20198K885 | List: |
| References | |
| PROC-860602, TAC-61635, NUDOCS 8606040159 | |
| Download: ML20198K899 (175) | |
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1 TENTATIVE & PRELIMINARY
- FOR DISCUSSION PU.RPOSES ONLY -
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4 SACRAM NTO MUNICIPAL UTILITY DISTRICT ACTION PLAN 3
i FOR
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PERFORMANCE IMPROVEMENT AT THE 1 RANCHO SECO NUCLEAR GENERATING STATION i JUNE-2, 1986 l
l TENTATIVE & PRELIMINARY
.- FOR O!SCUSSION PURPOSES ONLY -
t 8606040159 860530 PDR ADOCK 05000312 S PDR ,,
TENTATIVE & PRELIMINARY
- FOR DISCUSSION PURPOSES ONI.Y -
jq 1.0 Introducticn and Program Overview 2.0 Performance Improvement Actions Underway Prior to the 12/26/85 Event 3.0 Analysis of the 12/26/85 Event i
4 4.0 Management of the Restart and Performance Improvement Action Plan i
5.0 Restart and Performance Improvement Action Plan
- Systematic Assessment Programs
- Precursor Review Program
- Deterministic Failure Analysis
- B&W Owners Group Programs (Stop-Trip) 4
- 12/26/85 Event Response
- Plant Interviews
- System Review and Testing
- Programmatic Action Plans
- Plant Modifications
- Maintenance
- Lessons Learned l
- Training
- Operations and Procedures
- Quality and Quality Assurance
- Management Effectiveness a
TENTATIVE & PRELIMINARY
- FOR DISCUSSION PURPOSES ONLY -
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i TENTATIVE '& PREUt/dNARYl. a
- TGR DISCUSSION PURPOSES ONLY :.,.. .
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- Commitment Management
- Records and Data Base Managenent -
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- Configuration Management '
- Health Physics .,
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- Material Management 9
6.0 Restart Test Program ~ ',
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- l 7.0 Action Plan Schedule ,
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8.0 Transition (Later) 9.0 Response to NUREG 1195 Flydings
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APPENDIX A - District Board of Directors' Policy Statement on Performance Improvement at Rancho Seco.
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APPENDIX B - Comparison Table, Davis-Besse, TMI, Rancho Seco Restart Programs i
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TENTATIVE & PRELIMINARY
- FOR DISCUSSION PURPOSES ONLY -
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.I RANCHO SECO - RESTART, PLANT P AND MANAGEMENT IMPROVEMENT PROGRAM V)
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1.0 INTRODUCTION
& PROGRAM OVERVIEW The Sacramento Municipal Utility District (District) Board of Directors recognized in 1984 the need for an independent management assessment of the Rancho Seco Nuclear Generating Station operations.
This assessment resulted in the identification of several recommended management and programmatic changes. These changes formed the basis for the implementation of a performance improvement program by the District which was well underway at the time of the December 26, 1985 overcooling transient.
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Following the December 26, 1985 event, the District and the NRC independently conducted reviews to determine the nature and extent of contributions to the event from management, programmatic and hardware g deficiencies. The results of these reviews emphasized the need to N w.sg
( enhance the existing performance improvement program and to g accelerate its implementation. M h
The resulting Restart, Plant Performance, and Management Improvement Program (PP&MIP) described herein has been developed to assure the District's Board of Directors that Rancho Seco can be operated in a safe and reliable manner so as to impose no threat to the health and l
l safety of the general public, and at the same time, operate as a reliable and economic source of energy supporting the continued 1 Y economic viability of the Sacramento area.
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In support of these goa )$ am has these specific objectives.
- 1. A thorough understanding of the implications of the December 26, 1985 overcooling event using all resources available: NRC evaluations and audits, consultant studies, other utilities owning B&W plants, and the District's evaluations, root cause studies, and interviews with the Rancho Seco staff.
- 2. Identification and resolution of specific actions necessary to preclude similar events.
- 3. Recognition of the broader issues indicated by this event in the areas of management, training, maintenance, operations and the
/9 systems supporting these areas.
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- 4. Definition of corrective actions and programs that will significantly reduce the likelihood of reactor trips and challenges to safety systems, as well as assure that the plant t
will react nominally should a trip occur.
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l S. Establish rational criteria as to what actions must be completed l
before startup and what activities will comprise a longer-term improvement program.
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- 6. Definition of a testing program that will demonstrate the h
b material readiness of critical components and systems f p b
q_j operation. A8h p n r .-
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i) /}# $ ablish a deliberate and phased ascension to power program V
p % that will accommodate testing programs at appropriate power
' levels and establish confidence in the overall performance of the plant and its operators.
To efficiently and effectively accomplish these objectives the District has structured a program based on a management assessment of the programmatic and plant weaknesses, complemented by a detailed and systematic assessment of plant design features and industry experience.
The management assessment, conducted jointly by site management and a team from Management Analysis Company (a firm which has played a key role in many nuclear plant management, construction and operational improvement programs) consists of a review of current ongoing performance improvement program activities, documents associated with the December 26, 1985 event, previous performance assessments by consultants, INPO, NRC, and other outside agency assessments.
The comprehensive and systematic review process has been initiated and is well underway. This review is structured to identify specific actions required by the District to address relevant industry precursor experience, deterministic failure consequences analysis -
results, plant staff interview findings, and past recommendations from B&W Owners Group programs. This process is described %
detail in Section 5. %-% %
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gPhD-e t s from the systematic review process will be used, as they ome available, to adjust the program developed through the e management assessment process. While these adjustments are expected to be important, they are not expected to change the program scope or direction.
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.1 Performance Imorovement Areas The performance improvement areas developed through this process which form the key elements of the program are:
- 1. Plant performance improvement modifications: improve plant performance; improve operator plant performance monitoring; and keep the plant within the expected post-trip response envelope.
- 2. Maintenance program improvements: improve maintenance by applying systematic processes and criteria to the total program, as opposed to just class 1 equipment or systems.
- 3. Training improvements: expand systematic training programs beyond those provided for licensed operators.
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- 4. Operations and Operating Procedures: establish a clear commandandcontrolpolicywithintheoperatingcrewswhgl providing enhanced operating procedures which are \ @\
comprehensive and useful for their purpose. i g$
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,F4' 4 y Preparedness: implement policy, procedures, and 4 training which assure prompt and effective implementation of the Emergency Plan.
- 6. Quality: continue the development and implementation of a 4
Pro-Active Quality Program involved in all aspects of design, maintenance, operations, and training.
- 7. Management Issues: improve management planning, monitoring, and control systems; upgrade organization communication processes; and ensure that all management positions are staffed by personnel with the appropriate experience and qualifications.
. O U 8. Commitment Control: revise present system to provide accountability and control of process in order to reduce 1
backlog of commitments and improve credibility of commitments.
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- 9. Records Management: improve the records management systems including the records and Data Base content and accuracy.
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l 10. Configuration Management: expand existing program to [t M
I include the activities and needs of each of the Nucfle r l Departments in a consistent and systematic way b O
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r an Factors Engineering: integrate the process into the activities of each of the Nuclear Departments in a 4
e consistent and systematic way.
- 12. Health Physics and Radiological Controls: enhance the effectiveness of these programs by assuring that effective procedures, training, and process are systematically applied.
- 13. Materials Management: upgrade the material management system to assure it adequately supports maintenance and plant modification programs.
- 14. Lessons Learned Program: programmatically provide for the systematic evaluation of precursor events beyond the scope of the existing program operated by the Incident Analysis Group.
- 15. Other Programs: assure that the changing requirements and needs of the nuclear organization are uniquely resolved using systematic process.
.2 Criteria for Evaluation and Prioritization 4
The programmatic assessment of each improvement area leads to /
GY the development of items for resolution and disposition to g correct the identified deficiency, be it management or 4 g
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g4fr this purpose, criteria have been established to 4 4 ensure consistent prioritization of corrective actions and recommendations.
The criteria used to determine whether or not a recommendation
- should be implemented prior to restart are based on the impact
[ the recommendation would have, if implemented, on the frequency of reactor trips or the post-trip response of the plant.
, Specifically, these criteria are as follows:
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- 1. If implementation of the recommendation would significantly reduce the likelihood and/or frequency of a reactor trip or I challenge to safety systems, it should be considered for implementation.
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- 2. If implementation of the recommendation would significantly i enhance post-trip control, it should be implemented prior to restart.
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The above criteria are subjective and are not intended to require detailed analyses to quantify the benefits that may be derived from any particular modification. Rather, good e i
engineering judgment, based on previous operating experience
%Y 1 9 Rancho Seco and other B&W plants, is being used to evalu prioritize each recommendation. $
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\ Summarv Rancho Seco is an important energy resource to the District, the customer-owners of SMUD and the ratepayers of the northern California region. It is essential that Rancho Seco be returned to power operation and that it operate safely and reliably. The District is committed to taking those actions required to assure safe and reliable power operation prior to restarting Rancho Seco. The District is confident that the systematic processes described in this Action Plan will identify those items requiring implementation prior to restart. This process will also identify the performance and availability improvement items to be implemented in the long term to provide the community with an improved safe, reliable energy resource.
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Hith this submission of the District's Action Plan for Rancho Seco Restart and Performance Improvement, the Board, executive management, managers and staff of the District affirm their commitment and resolve to achieve excellence in the restart, operation and maintenance of the Rancho Seco Nuclear Generating Station. He will continue aggressive and responsible pursuit of
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that commitment, including providing the necessary resources.
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l As the District management team shaped this program, it became clear that there was need for policy direction regarding the s% s g' l long-termfutureoftheRanchoSecoNuclearGeneratingStatign9 q@
This policy direction is considered essential to pr @
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p\g ndation for long-term planning. The policy direction has i
4- $ been prepared and unanimously endorsed by the Board of p Directors. A copy of this Board Policy Statement is included in Appendix A. ;
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4 59 O 2.0 PERFORMANCE IMPROVEMENTS M RIOR TO THE 12/26 EVENT U gyv In late 1984, the Board of Directors of the Sacramento Municipal
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Utility District recognized that the District had a significant i number of challenges facing them. Foremost among those challenges i
- 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 performance of the Board, the Staff, and plant i personnel.
l 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 including plant modifications, personnel performance, management systems, analytical capability, training, and organizational
- 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.
l 1 During 1985, in recognition of the above situation, the Board embarked on an overall program to upgrade the District's organization
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) and operations thereby improving the effectiveness and reliability of @$
plant performance. The thrust of the improvement program wa t 4e -
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Y, with a large spectrum of mt.nagement and organi $ ssues including:
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- 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 f
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. Establishment of a clearly defined maintenance program.
. Establishment of an effective systematic troubleshooting program
. 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 9 these plans a number of activities were initiated, beginning ibh 8 Spring of 1985, and the program continued through the sp
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Q b d ptember of 1985, Retired Major General Dewey Lowe joined the V $ g%gs4 4 District as the General Manager reporting to the Board of Directors.
1 This move represented a substantial step in implementation of the Board's program by infusing new expertise and direction into the
- organization at the very highest level.
2.1 Projects Underway Prior to December 26, 1985 2.1.1 Staffino and Orcanization Significant to the conditions determined in the 1984 study was the recognition of the sizable growth in plant staff, without a corresponding restructuring or expansion of the management staff. A new organizational structure was approved with six Nuclear Departments reporting to the AGM-Nuclear.:
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- 1. Nuclear Operations
- 2. Nuclear Engineering
- 3. Quality
- 4. Nuclear Training i
- 5. Nuclear Licensing
- 6. Nuclear Projects ef$
The last three departments listed had previously been elements wink the Nuclear Operations and Engineering Departments. gb g@
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. d approved this structure, and to provide the staff to fill 4 ew management and technical positions, nationwide recruiting efforts were mounted. All new key positions were staffed by early 1986.
Numerous other structural changes occurred within the various departments. The purpose and effect has been to reduce the diverse managerial requirements upon individual supervisors and superintendents by establishing new divisions and alignments which bring similar functions under the direction of a single manager, i
- Hiddle management is now better able to cope with the demands of the 4 groups for whom they are responsible. He are seeing them spend more tiene on details while interacting with the personnel and projects I
coming under their purview. As a number of these people have been at i
O Rancho Seco for less than a year, there has been a considerable 1
N.) 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 ,.
l 1s receptive a the programmatic approach and conmitted to
- attention-to-detail and accountability.
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2.1.2 TRAINING PROGRAM y 2.1.2.1 Mana Msructure Previously, the 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 departmental status to ensure top management involvement.
As of June 1985, the training organization became a department answering directly to the Assistant General Manager, Nuclear. The position of Training Manager was established and filled from outside
,O the District organization, bringing a new perspective and experience D 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
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the'trainina department with only one exception. This exception is the Fire Brigade training program which will also be transferred 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 5.B.3, will result in management recognition of the training function as an eJM 9
integral part of plant operations and ensure effective coordin
- b. of the training function with all other nuclear organi V functions. '
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l 2.1.2.2 ObA itation Effort The District has been committed to INP0 Accreditation for its '
training programs for the past several years in an effort to improve the overall training program. The District is using a phased
.i approach for this effort. To focus on obtaining accreditation for j various department functions on a sequential basis, starting with 3
operations.
The first phase, consisting of Senior Reactor Operator (SRO), Reactor Operator (RO), Shift Technical Advisor (STA), and Non Licensed Operator (NLO), received accreditation in April 1986. The remaining six training programs, which involve maintenance training, chemistry and radiological protection technician training, and technical
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V) support staff and managers training programs will be submitted for accreditation in July 1986. The accreditation process will ensure that the shortcomings identified in maintenance training are corrected.
I 2.1.3 Maintenance Procram f
Prior to the December 26, 1985 event, the District had initiated a number of actions designed to improve its maintenance program. In addition, at the time of the event, several specific maintenance program enhancement actions were underway. These included: '
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the Maintenance Manager position. g(c, 2-6
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. staffing levels authorized for maintenance in the 1986 gr,c dget. This included allowances for Preventive Maintenance Supervisors and dedicated Preventive Maintenance Crews in the
. Mechanical, Electrical, and I&C Groups.
. Dedicated Planning Personnel authorized for the Electrical, Mechanical, and I&C Maintenance Groups.
. Formation of a centralized scheduling group to provide overall prioritization and integration of maintenance work with other plant activities.
. A full-time consultant was reviewing and refining the Preventive Maintenance Program.
2.1.4 Systematic Troubleshootino In March 1985, Rancho Seco was shutdown for a scheduled 90-day refueling for cycle 7. The plant was returning to service 94 days later when a RCS vent line cracked, causing a shutdown due to excessive 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 -
t occurrence, a special systematic troubleshooting program (based upon 9
the Davis-Besse NUREG 1154 Appendix B methods and criteria) was b implemented. This was in addition to programs that were n
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a Root Cause Analysis, the support work done by tne B&W wners Group Transient Analysis Program (TAP) team, and the Nuclear 4
' Operations Trip Report investigation.
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.
2.1.5 Root Cause Procram 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 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 programm (fx g determinations. 4 $
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3.0 ApNs THE 12/26/85 EVENT v p On December 26, 1985, while operating at 76 percent power, the Rancho Seco Nuclear Generating Station experienced a loss of DC power to the Integrated Control system (ICS). Subsequently, the reactor tripped on high reactor coolant system (RCS) pressure followed by a rapid reduction in RCS temperature and pressure (caused by ICS controlled valves failing to an open position) and automatic initiation of the Safety Features Actuation System (SFAS).
The District's immediate response was to institute systematic troubleshooting and investigations into the event while proceeding to take the plant to cold shutdown. NRC Region V dispatched an Augmented O Investigation Team to the site, to perform their own assessment of the
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'/ event. Later, this team - with some personnel and charter changes -
became the Incident Investigation Team (IIT). The IIT published their findings as NUREG 1)95 on February 24, 1986. That report, and the District's own investigations, agreed on the direct causes of the event and the analysis of the event consequences on plant equipment and the environment. NUREG 1195 expanded significantly the investigation of precursors to the event. The report stated that the District's history of responsiveness was characterized as being " narrowly focused". The introduction of these " retrospective" issues expanded the scope of necessary event analysis to look beyond the event itself and seek out gg .h' the prcgrammatic issues which allowed it to occur. That effort i$.i [
discussedinSection4,whilethissectionaddressestg g
/ investigations done relative to the event. M N])
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b O 3.1 Event Investication Root Ca4N as b %}$
F As a result of this transient, a comprehensive root cause investigation was initiated by the District's Incident Analysis Group. The investigation considered personnel, equipment, procedural, training, and managerial issues related to the event. Due to the complexity of the event, the Incident Analysis Group considered the event as a composite of five (5) major event themes:
- Loss of ICS power
- Rapid cooldown
- Makeup pump damage
. Health physics concerns O + Emergency plan
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It was concluded that the most significant of these themes was the rapid cooldown of the RCS and that the root cause was a failure to implement design changes which would compensate for known design weaknesses. In addition, compensatory measures to compensate for design deficiencies, such as training and procedural guidance, were not implemented. -
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The root cause and contributory causes of the five event themes s#
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S 93 Event Theme gM u Contributory Cause loss of ICS Power Manufacturing defect.
Rapid Cooldown Failure of District Inadequate Procedures management to evaluate the Inadequate Training full scope of and act 1 expeditiously on recommend-dations/ requirements of industry groups / regulators.
Makeup Pump Damage Inadequate Procedures. Personnel Error Health Physics Lack of clearly defined Procedural Non-Concerns division of responsibili- compliance ties between the operations personnel and health physics personnel in areas involving health physics.
Emergency Plan Procedural Non-compliance Inadequate Training 3.2 Related Investication and Analyses In addition to the root cause investigation conducted by the Incident 9 0 o -----p- -
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b us other investigations and analyses which focused on specific
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g spects of the event. Some of the more significant items are tabulated
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. Equipment Failure Investigations
. ICS
. Makeup Pump
. Radiation Monitor i
. Effects of Overcooling Transient on Components
. Operational Review (including adequacy of procedures)
. Human Factors Evaluation
. Adequacy of Training
. Thermal / Hydraulic Response of Reactor Coolant System 4
. Health Physics
. Emergency Preparedness In many cases, these evaluations supported and contributed to the root cause investigations and were a source of many recommendations for f
improving the performance of Rancho Seco. To date, nearly 400 recommendations have been identified. An action list and tracking system has been developed to ensure that these recommendations are g' properly considered and dispositioned. TheserecommendationsarebeigS processed in accordance with the methodology discussed in Se % c)N lO 1
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I 3.3 nM ons in Resconse to NUREG-1195 Findinas vp-The District has taken a three part approach to NUREG-1195 to ensure that the concerns and issues raised are properly addressed. The District's approach includes:
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. a review of NUREG-1195 for technical accuracy,
. a review of NUREG-1195 for plant improvement recommendations, and
. response to the NUREG-1195 findings and conclusions.
1 The District provided the results of the review for technical accuracy t
to the NRC by letter dated April 24,1986 (Mr. R. J. Rodriquez to Mr. F.
J. Miraglia).
iO The District h't completed the review of NUREG-1195 for plant 1"arovement recommendations. The resultant recommendations are being addressed as a part of the overall improvement program.
i The District's responses to the findings and conclusions of NUREG-1195
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are provided in Appendix 9.0. A number of the findings and conclusion 9 .
b are being addressed as a part of the Plant Perforamnce Improve g Program, Section 9.0. 4 1
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4.0 MANAGEMENT OF THE RESTART AN CE IMPROVEMENT PLAN This section describes the various management processes and organizational relationships associated with the implementation of the Restart and Performance Improvement Plan.
Sucolemental R9 source.1 In addition to the norn.a1 resources associated with plant operations, the District has:
a) Established two independent advisory groups reporting to the General manager.
' O b) Retained the services of Management Analysis Company to support the development and implementation of the Restart and Performance Improveuent Plan.
l The two independent advisory groups consist of an Improvement Program I
Overview Panel made up of Admiral E. P. Hilkinson (Ret.), Dr. T.
Pigford, and M. Rowden, and An Independent Review group made up of Messrs. J. Jackson, R. De Young, J. O'Hanlon, and A. Gehr.
l These two groups have been established to provide the General Manager with an assessment (based on their significant experience in the gp '
commercial nuclear industry) as to the effectiveness of the improv program and the readiness of the plant to operate. g 1-1
Ngement Analysis Company (MAC), a firm which has played a key role in many nuclear plant management, construction and operational improvement programs, has been retained to assist SMUD in the devetapment and implementation of the Plant Performance and Management Improvement Program. MAC has provided an experienced nuclear executive to act as Assistant General Manager (Nuclear) and provide overall direction. In addition, a MAC team of experienced nuclear plant management, engineering, maintenance, training, information systems specialists, licensing, quality assurance and organizational effectiveness professionals will:
- 1. Review the PP&MIP and assist in the development of a detailed action i
plan outlining actions required to be completed before restart, the p
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essential elements of a comprehensive component and systems testing N program, and definition of a longer termed action plan which will, over the years, result in achieving the District's overall goal of excellence in nuclear operations.
- 2. Once the detailed action plan for restart and the basic program elements of the PP&MIP have been accepted by the District and the NRC as the basis for restarting Rancho Seco, MAC managers will be placed within the nuclear organization as deputies to selected key i SMUD managers. They will assist in implementation of the PP&MIP, providing the SMUD managers with the benefits of their experience in 6Y-similar improvements programs. Whenthisassistanceisnolongeg deemed useful, the MAC role will be terminated. %
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( Oraanization Funct The organization and functions will change as the implementation of the Restart and Performance Improvement Program progresses.
Restart Oraanizat'on During the first phase (Restart), the organization will be as shown on figure 4-1 (later). This figure shows a restart manager reporting to the AGM with responsibility for coordinating and managing the implementation of all restart actions. During this period, normal day to day activities, which are not part of the restart program will be handled by the line managers through their normal lines of authority.
Post Restart Proaram Following the completion of the power ascension testing, the restart manager will change focus to the implementation and closeout of the performance improvement action plans which are identified as being initiated prior to restart and implemented in a timely manner.
The two independent advisory groups will continue to assess the progress
, and effectiveness of the performance improvement programs until the i
action pians identified as being initiated prior to restart and those action plans initiated prior to restart and implemented in a timely m_a,e_p,etedo,t.0,ea,s,_,esta,t..h,c, eve,_s l0 l es s -
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RY IMINA Y-ONI.
PRELOSES
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%g Ytf"Ity % TIVE NP
/bN % SCUSSIO DI FOR The restart and performance improvement program defined in this document is structured to accomplish the overall objectives of the program identified in the introduction and program overview section. With the concurrence of the NRC these objectives will form the basis by which restart program activity adjustments will be assessed and dispositioned. This is an important feature since it is the composite of all the various program activities which satisfy the objective, and not necessarily any single given activity. -
Plant Performance Imorovement Procram (PPIP) )
The PPIP process is defined in detail in the implementing procedure QCI-12, the Plant Performan e and Management Implement Program. The purpose of this process is to ensure that potentially meaningful conclusions or recommendations from the unique input programs nd industry sources, which may not have been comprehensively addressed i earlier, will be systematically reviewed and reconsidered for their
- significance and applicability to Rancho Seco. This program will also include new detailed evaluations for systems which have a high potential for impact on future plant performance.
A flow chart of the PPIP process is provided in Figure 4-1. The j
guidance for the individual organizations discharging their ,
responsibility under this program is contained in QCI-12. In general, a [
review and evaluation is conducted with recommended actions c [
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FIGURE 4-1 g PERFORMANCE IM OGRAM INPUT Selected Deterministic Projects Failure Consequences Precursor BH0G Stop-Trip Review Program 12/26 Event &
Plant Staff NUREG 1195 Interviews Action List EVALUATION Recommendations Review & Resolution Board DISPOSITION Disposition by Feed back Performance Programmatic to RRRB Analysis Group Trend Analysis l l CLOSE, ACTION No further ACTION Non-Startup to action required required for startup Living Schedule or CCL I
IMPLEMENTATION Implementation using approved ECN/DCH, Operations, t Power Ascension Procedures, Engineering Training Programs epd Test Program M
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O hY e Re:ocaendation, Review, and Resolution Board (RRRB), which is
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,\ made up of senior representatives of the various line organizations.
-The RRRB reviews the recommendations on its technical merits and recommends the appropriate disposition to management (PAG)
The Performance Analysis Group is 'made up of the Nuclear Department managers reporting directly to the AGM-Nuclear. This group determines the final disposition of the item and deternines whether or not the
, existing programmatic actions must be adjusted or not to satisfy the basic criteria of section l.0. Responsibility for implementation of appropriate actions to close the recomended item is assigned as part of this PAG process.
O On a quarterly basis, the AGM-Nuclear will meet with the Improvement Program Overview Panel to review in general, progress of the restart and performance improvement plan.
On an informal basis, the AGM-Nuclear will stand ready to provide a progress report of the Restart and Performance Improvement Plan with any local, state, national, or industry group deemed relevant to this project by the GM and AGM-Nuclear.
Action Plan Trackina. Recortina. and Close-Out Audit The Restart and Performance Improvement Action Plan items are entered on b g o --
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- N b am (Project II). The status of each item will be updated on a
,@ weekly basis during the period of program initiation to prior to restart and a monthly basis following th completion of power ascension tests.
This status report shall be provided to the AGM, the Restart Manager, the Program Scheduler, the independent oversite groups, and the responsible organizations. The reports shall identify which of the following status conditions each activity exists in:
a) progress toward completion (key milestones and 1. complete).
b) Complete awaiting QA closure.
c) QA closure complete awaiting NRC closure, d) NRC closure complete or not required.
e) Entered on living schedule or CLL.
O V The living schedule will be used where long term items are identified.
Action Plan Closeout Verification A Quality Assurance Verification Audit will be conducted to close out each element of the performance improvement plan associated with actions to be completed prior to restart and items to be initiated prior to restart and completed in a timely manner. This audit can apply sampling techniques where appropriate but will be of sufficient depth to assur that the objectives of the improvement program are met.
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The AGM-Nuclear will provide status reports to internal and external groups on a regular basis.
Internal On a monthly basis, the AGM-Nuclear will meet with the Rancho Seco Implementation Committee of the Board of Directors to review in detail, progress of the restart and performance improvement plan. At the subsequent full Board meeting, he will present an overview of his report.
On a monthly basis, the AGM-Nuclear will meet with the General Manager,
) Assistant General Managers and relevant staff to review in detail Q 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 status report for District employees.
! External l
l On a monthly basis, the AGM-Nuclear will meet with NRC to review in l
detail, progress of the Restart and Performance Improvement Plan.
On a monthly basis, the AGM-Nuclear will meet with the Independent b
Review Grm n to review in detail, progress of the restart an I (w/n) i i
performance improvement plan. 4S
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5.0 RESTART AND PERFORMAN N This section of the program plan describes the Systematic Assessment Programs (Section SA) used to identify those items which are Ilkely precursors to possible future transients, and the programmatic actions (Sectin 58) identify the implementation actions to modify the plant design and improve plant programs.
i The recommendations generated during the Systematic Assessment Reviews and approval by the Performance Analysis group will be input to the programmatic action plans for implementation. These programmatic action plans will be adjusted as required to accommodate the new items.
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5A.0 D f3 SYSTEMATIC ASSESSM q g %
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Ihe Plant Performance Improvement Program (PPIP) portion of the PP&MIP was structured to accomplish a systematic assessment of the physical plant, its operating procedures and interfaces with training, maintenance. It is specifically looking at documents and industry information for precursor event and findings which may have not been incorporated into Rancho Seco due to past " Narrow-Focus" attitudes and practices. In a similar vein, it is aggressively looking to the facility staff for input related to concerns and issues which they feel need to be programmatically or specifically resolved.
The program also provides for a comprehensive Deterministic Failure O Consequences review of the powered equipment (air or electrically powered) and control systems. This effort anticipates the untimely failure of both primary and secondary systems, or equipment, and is to assure that appropriate procedures, training, or equipment is in I place to preclude adverse post-trip plant response.
Systems Review and Testing is being done to demonstrate the material I
condition of the plant in both the pre-startup and power escalation modes.
/se' The program to manage these input processes is implement QCI-12. Not only does it systematically develoo
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lons, but provides for the systematic evaluation, b, position, implementation and prioritization of the numerous 1
, @ recommendations. The balance of this section discusses the input
- areas and process in detail. It should be noted that as a i
consequence of the December 26, 1985 Event, Systematic ,
Troubleshooting and Root Cause projects have developed over 300 i
! recommendations which are also being dispositioned by this program. I 2
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s 5A.1 PRECURSOR REVIEW P
,1Y Objective--The objectives of the Precursor Review Program are to use a systematic review process to identify events or conditions that have previously occurred and to determine their significance to Rancho Seco. From the events and conditions that are judged to be applicable and significant to Rancho Seco, a specific recommendation will be made to improve the affected plant area (design, operations, maintenance, etc.) to either preclude the occurrence or minimize the effect of the event or condition at Rancho Seco. The identified problems and improvement recommendations will be input to the plant performance improvement process for disposition.
n Scope of Work--The scope of work to be performed by the Precursor Review Program is divided into two parts.
I A. 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:
All Transient Assessment Program (TAP) Category C ,
t transients will be evaluated and investigated f ~
b applicability and impact on Rancho \ g@
5-4 i _. _- _-_. _ _ _ , _
ory B TAP events will be reviewed to determine if gb $YnyoftherecommendationsmadeareapplicabletoRancho
', Seco and to determino 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.
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
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4 V described above will be completed before plant restart.
B. Other Document Reviews In addition to the review of the TAP data, the following other documents will be reviewed by a multi-discipline experienced team:
(1) Licensee Event Reports and Occurrence Description Reports O
l (2) Significant Operating Experience Reports (50ER) theInstituteofNuclearPowerOperationg 5-5
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(3) issued by the NRC Office of Inspection and :
b #8hnforcement (4) Notices / Circulars issued by the NRC Office of Inspection and Enforcement (5) Babcock and Wilcox Reports (Preliminary Safety Concerns, Site Instructions, and other applicable BAW reports)
This review will be completed prior to startup.
This program provides for a reverse chronological review starting from 1985. Prior to start-up, documents dating back to March 1978 will be reviewed.
Criteria and Methodology for Precursor Evaluation Each document will be reviewed to determine whether issues are applicable to Rancho Seco. For each document a Precursor Review Checklist will be completed. For those issues which require a recommendation, the Precursor Review Recommendation Form will be completed. The results will be forwarded to the RRRB.
Schedule--All evaluations of all identified items will be before plant startup. 4 6 s'
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5A.2 DETERMINISTIC FAILURE NALY'iIS o -
Purpose--The objective of the Deterministic Failure Consequence Analysis project is to determir;e the consequences of failures of systems on power operation or post-trip response capability, and to evaluate procedural guidance available 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.
.O
() 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 effects of most postulated plant failures. The analysis will identify affected systems which challenge or adversely effect the capability to mitigate transient condi tions .
Methodology--Each system will be reviewed as described below:
Loss of Electrical Power--Teams will analyze each 480V bus, its source and loads. Each team will review electrical elementaries [
beginning at the end loads. EachbreakerofftheMotorContrg19
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5-7
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bl an assumption of no physical positioning at circuit (V) .
g ers, 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 loss of the individual loads are evaluated, the loss of the source 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 sources.
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 I
will be analyzed.
Electrical elementary drawings shall be " yellow lined," identifying the breakers " opened" and affected components to ensure each load is addressed.
Loss of Instrument Air An evaluation of the loss of instrument air event will be performed. Individual components on the Instrument Air P&ID will be " failed" and the effect upon the Instrument Air System and the plant analyzed. The entire system will then be " failed" and the effect on the plant analyzed. Individual instrument air headers [
will be failed, components affected by this failure listed, and tb C' effect on the plant analyzed. The P& ids will be " yell g6
% ensure that each component is addressed. I g%
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.lcob CS/NNI--The loss of ICS and NNI will be evaluated by g e endently falling each system (ICS or NNI) parameter, identifying
/ the affected components in other systems, and determining the effects upon the plant. Aopropriate drawings will be " yellow lined" to ensure each component or parameter is addressed.
Recommendations developed during the reviews 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 system engineer assigned to that system.
Schedule--This project, including the evaluation of recommendations generated from the review, will be completed before plant startup.
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5A.3 B&W OWNERS GROUP PROGRAMS -
Objective The objective of the B&W Owners Group (B&WOG) 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 when they occur. The specific goals of the programs are:
By the end of 1990, the average plant trip frequency will be less than 2 per year.
By the end of 1990, the number of complex transients as classified by measurable parameters (Category C as defined by the B&W Owners Group) will be reduced to less than 0.1 per year 5 based on a moving three-year average.
The District will fully participate in the STOP-TRIP Program.
i 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.
i Program Scoce The STOP-TRIP Program is similar in many ways to the District's Plant Performance Improvement Program. However, it is designed to allow all utilities with B&W-designed plants to provide input and evaluations. The program is an extension and expans \b from previous B&W Owners Group activities aimed ge98 5-10
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Q W reactor trips which occur. Thus, a number of programs are 4 [eadyunderwayandrecommendationsbeingpreparedforevaluation
' 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:
- 1. Independent Sensitivity Study
- 2. Risk Assessment Review
- 3. Operating Experience Review i
- 4. Operator / Maintenance Reviews
- 5. System and Component Reviews b 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 Group submittals to the NRC.
i All B&W Owners Group reports issued since 1980 will be reviewed for recommendations applicable to Rancho Seco to assure that past recommendations have been considered. The recommendations will be evaluated as to their appilcability to Rancho Seco as well as whether they have already been implemented or not. Any that are applicable will be forwarded to the RRRB for implementation / evaluation.
Methodologv/ Procedure--The District's STOP-TRIP Program Co ihat
! N' (STPC) is responsible for providing the interface be t
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\@ ON ly and the B&W Owners Group Programs. The STPC shall review d recommendations of these programs and compare them to the
@ recommendations of the PPIP. Recommendations that are not in the PPIP shall be submltted to the Recommendation Review and Resolution Board (RRRB) for evaluation. The evaluation shall determine whether the recommendation becomes a restart restraint or a long-term item for future resolution.
Schedule-Time of Performance ,
In recognition of the fact that the B&W Owners STOP-TRIP program is a relatively long-term project, any items issued by the STOP-TRIP program after startup will be addressed through the District's Reliability Engineering Program.
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. ,p 5A.4 PLANT Objective and Purpose - The purpose of the interview program is to surface previously unresolved, but "known" problems which can (1) cause reactor trips and contribute to the severity of transients, and (2) degrade plant reliability or the optimal performance of the operating personnel.
Project Scoce - This program will interview personnel from each plant organizational group. These persons will be encouraged to identify system, component, or operational problems or concerns which they are aware of 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 x functional group and the interview coordinator will select personnel for interview if insufficient volunteers come forward.
Interview Process Methodology - The interview program is to cover essentially a cross section of all plant personnel and is intended to
, encourage those personnel to identify issues or concerns which they j are aware of that, when resolved, can contribute to optimal and reliable operation. This program will be coordinated by the l
Interview Program Coordinator and will be handled as part of the ongoing Nuclear Projects Department activities. Each interview ed '
S9 @N be documented using the interview form. $\4 g
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on and question form was prepared and is presented to g ts Y g interviewee prior to the interview. Each interviewee is asked 4
- for information about his/her background and is brie':3 as to the purpose of the interviews. The questions on the forms are then discussed one-by-one. Each interviewee is asked to expand on each answer until the interviewers feel no further meaningful information is available.
The Interview Project Coordinator consolidates the compiled list of concerns / recommendations. He will forward the recommendations to the RRRB using the Recommendation / Resolution sheets. The Interview Program Coordinator shall assure that the concerns and
! recommendations have been acted upon and dispositioned to the RRRB.
/ Schedule - The program interviews, including evaluation and g' dispositionoftherecommendationswillcompletedprirtA8 restart. p 4
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F-5A.5 5 k D ST NG Several plant systems play an integral role in maintaining good plant control during power operation and/or are vital to post-trip control. The District has utilized this criteria to select systems for review.
Objectives - The objective of the system review program is to systematically evaluate these systems selected to ensure that:
The systems as installed and its design basis are adequate to support plant operation under normal and upset conditions as appropriate.
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- Engineering and procurement is initiated on modifications for Installations scheduled after start up.
l Long-term plans for resolution of minor problems are initiated.
Scope of Work - This review takes into account the system functions and the systems capability to perform these functions during plant startup, and transients. This review addresses both District and industry known design deficiencies, operational problems and maintenance problems.
Engineering items identified to date are listed in Section SB and $'& g '
i have been assigned approcriate priorities. Section6.0identgiesge.
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O g program that will be utilized to verify the operational 4 abilities of these systems. The testing section covers safety systems and non-safety systems that are identified as potential reactor trip initiators or could contribute to post-trip control problems.
Systems - The following systems have been identified to date and l reviews have been initiated.
i' ICS/NNI Auxiliary Feedwater Component Cooling Water Control Room /TSC HVAC s Diesel Generators l
Startup Transformer #2 Main Steam Line Failure Logic 125 VDC E & F Busses I
- 120 VAC Instrument Air As part of the systems review, the District will review the commitments made relative to NUREG-0737 and Reg Guide 1.97 to verify that the intent of each commitment was satisfied. In addition, a review of the status of those items which are not yet completed to will be highlighted to provide a basis for required management 1 4' D'
b-attention to ensure future commitments will be met. Unccm g
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i ek be evaluated for restart restraints against the restart 4 eria described in Section 1.
s' In addition to the techniques and processes that will be used to i
review the selected systems, the ICS/NNI, instrument air, and electrical distribution systems will be examined using an additional technique called the deterministic failure consequence method. This method is described in Section SA.2.
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j 58.0 PROGRAMMATICACTIONEL@lku i This section lifentifies the specific programs, and program enhancements, which are being implemented to resolve the lessons-learned, recommendations, and programmatic deficiencies i
identified by the Action Plan. Each subject area is discussed in l general, followed by specific commitments relative to that issue.
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m 58.1.1 PERFORMANCE IMPROVEMENT AREA: Ig ONTROL SYSTEM (ICS) AND INTERFACING SYSTEMS b NfM s
The Integrated Control System is an important element in the control of the plant during normal operating evolutions. 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 following failures of the ICS. The general programmatic actions and specific 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, reduce the potential for reactor trip and improve the operators ability to maintain the plant within the post trip window, are being developed or have been developed by various organizations.
- l The objective of this general programmatic action is to assemble, review, and implement those changes which are judged g '
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.1 Agt e Comoleted Prior to Restart p h@
- 1. Assemble and review the various source documents containing recommendations relating to reliability performance improvements of the ICS and interfacing systems.
- 2. Identify those recommended actions which have, and have not, been addressed by modification of equipment and procedures at Rancho Seco.
- 3. Develop and apply an evaluation criteria for those items which have not been effectively addressed to screen and prioritize these recommendations.
r l 4. Where appropriate, develop and implement design changes.
.2 Actions to be Initiated Prior to Restart and Completed on a Timely Basis
(
- 1. Develop and implement those design changes or ,
enhancements that were not required to be impleme b
prior to restart, consistent with their as priority. 4 O
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.3 Long-Term Implemental s or Actions of a Continuing Nature h-
- 1. Actively participate in the B&W Owners Group efforts to upgrade and replace the Integrated Control System.
1.b Actions to Reduce the Impact of Power Loss on ADV's, TBV's, ar' AFW Control Valves The loss of power to the ICS during the 12/26/85 event resulted in the ADV's, TBV's, and AFW control valves falling to mid-stroke due to the bi-polar nature of the ICS. This resulted in an overcooling transient following the reactor trip.
,V 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.
.1 Actions to be Completed Prior to Restart
- 1. Provide a control station in the Control Room, from which the operator can manually control the AFH valves independent of the ICS, and which will position theps\
valves to a predetermined (minimum) position gr? p @ "
detection of loss of ICS (DC) power b Ov 5-21 i
CN gqt2. vide controls in the Control Room from which the 5% \ operator can coerate 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 i was identified following the 12/26/85 event as a potential i
single failure point which could contribute to the loss of ICS.
The objective of these actions is to evaluate the potential impact of the power supply monitor and to implement charges l
l which will effectively improve the single failure likelihood.
.1 Actions to be Completed Prior to Restart l'
- l. Determine the contribution ne Power Supply Monitor had in 12/26/85 transient.
- 2. Evaluate the potential improvement to ICS reliability l
if redundant PSM's installed.
- 3. Determine the potential benefits to be cbtained through the installation of redundant PSM's. ,
- 4. Develop designs for redundant er indepen J .
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x $ 6#Ab to Improve the Status Indication on Loss of ICS Power st #
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 loss of DC power.
The objective of these actions is to implement design changes which are effective in addressing this deficiency. To accomplish this objective the following actions will be taken.
.1 Actions to be Completed Prior to Restart i
fp 1. Provide separate windows for the status indication of the following:
1
- ICS Trouble (fan failure, power supply failure).
ICS Failure (Loss of DC Buss).
l 1.e Actions to Evaluate the Failure Consequences of Various ICS Inputs. Outputs, and Components The ICS has undergone a number of changes since a generic pN, failure / consequence analysis was performed in (approximately9 S 1980, BAW1564. These changes include not only physic 4fl ug%
philosophical design basis changes. $6 6 G
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d' m ve of these actions is to update and evaluate the 4 act the changes have had to the overall functional performance of the ICS relative to individual failures of various ICS inputs, outputs, and components. To accomplish this objective, the following actions will be taken.
.1 Actions to be Initiated Prior to Restart and Completed in a Timelv Manner
- Participate in B&W Owners Gro*1p efforts to perform a generic failure / consequence evaluation of the Model 820 ICS. If the BWO6 FMEA is not initiated by August 1, perform an independent, Rancho Seco specific FMEA.
d
.2 Long-Term Implementation Actions or Actions of a Continuing Nature
- Evaluate results and applicability of B&W Owners Group evaluation findings and recommendations to Rancho Seco.
- Conduct supplemental evaluations as required to ,
achieve Rancho Seco specific information.
b
- Develop and implement applicable modific l Rancho Seco ICS. .
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1.f MN$nIst dress the Adequacy of ICS Power Supply I h %
During the 12/26/85 transient, ICS DC power was lost, then recovered. During the power outage, the Hand / Auto stations were inoperable.
The objective of the:e actions is to evaluate, develop, and
^
implement beneficial design changes to address this design weakness. To accomplish this objective, the following actions will be taken.
.1 Long-Term Actions or Actions of a Continuing Nature
- Evaluate need for battery backup based on reliability 3
of power supplies, recent modifications, etc.
- Evaluate bus backup.
- Evaluate Hand / Auto station backup.
- Evaluate alternative backup methods based on reliability of power supplies and recent I
modifications. The alternatives will include; backup i
( batteries, DC Bus backup, and Hand / Auto backu QN Develop and implement applicable modip'!dpi O
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58.1.2 PERFORMA AREA: NON-NUCLEAR INSTRUMENTATION (NNI)
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 during normal plant evolutions. 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 minimize reactor trips and maintain the plant within the post trip window, are being developed or have been developed by various organizations.
The objective of this general orogrammatic action is to assemble, review, and implement those changes which are judged to be safety or cperationally beneficial. To accomplish this objective, the following actions will be taken:
$b
.1 ActionstobeInitiatedPriortoRestartandComolste . $1 Timely Basis 9 9 o
a 5-26
and review the various source documents
{.9 V b c$DY containing recommendations relating to reliable
, performance of the NNI.
4
- 2. Identify those recommended actions which have and have not been addressed through modifications to equipment and procedures at Rancho Seco.
- 3. Develop and apply a screening and prioritization evaluation criteria for those recommendations which have not been effectively addressed at Rancho Seco.
- 4. Develop and implement the appropriate design changes based on their established priority.
.2 Long-Term Implementation Actions or Actions of a Continuing Nature t'
- l. Actlyely participate in the B&W Owners Group efforts to upgrade or replace the Non-Nuclear Instrumentation equipment.
2.0 Acticns to Address the Adequacy of NNI Power Monitors As a result of the 12/26/85 event, the ICS power supply monitor (PSM) was identified as a potential single point which coul g$bW p cause a loss of ICS DC power. The NNI has the sam p th DN 5-27
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b p g p.Trering configuration as the ICS. The objective of these I6 acticne is to evaluate the potential impact of these power supply monitors and to implement changes which will effectively improve the single failure likelihood or develop.
.1 Actions to be Completed Prior to Restart
- 1. Determine the contribution the Power Supply Monitor had in 12/26/85 transient.
- 2. Evaluate the potential improvement to NNI reliability if redundant PSM's installed.
. 3. Determine the potential benefits to be obtained C through the installation of redundant PSM's.
- 4. Develop designs for redundant or independent PSM's.
2.c Actions to Address the Adequacy of Status Indication for NNI, and Affected Instrumentation on Loss of NNI Power On loss of NNI-X or NNI-Y power, certain Csecrol Room indication and control is lost. This issue concerns (1) identification of the NNI status - e.g., that NNI-X or NNI-Y have been lost, and (2) identification of Control Room instrumentation associated with NNI-X or NNI-Y. (Presently, on loss of either, the g e
operator is directed by procedure to alternate (non-fiff)?
instrumentation.) $9. O 5-28
pB6 S b N je tive of these actions is to effectively address this x ,1 plant weakness.
.1 Actions to be Completed Prior to Restart
- 1. Identify instrument loops associated with NNI-X or -Y.
- 2. Evaluate need/ feasibility of identifying power source for Control Room instruments.
- 3. Initiate design changes to provide Control Room instrumentation identification.
' 4. Provide separate annunciator windows to indicate:
Loss of NNI-X (DC)
- Loss of NNI-Y (DC)
- Loss of NNI-Z (DC, swltching supply)
! - NNI trouble (fan failure, single power su 9
failure) 996 i
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?Yq O 58.1.3 PERFORMANCE IMP EEDWATER AND STEAM SYSTEMS v -
The design features and performance 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 to Improve Emergency Feedwater Initiation and Control Rancho Seco developed a design and established a schedule for the implementation of an Emergency Feedwater Initiation and
'(g/ )
Control System (EFIC) in response to industry lessons learned.
This system provides:
a) Safety Grade Auxiliary Feedwater initiation and control.
i l
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 the cycle 8 and cycle 9 refueling outage, with the g\b i o s,, _ , _ .so,..s,st.mh.,,,,,.c. ,> ,,,
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(,hh would have reduced the severity of the cooldown ansient of 12/26/85.
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 i control of ADV's, TBV's, and AFH control valves, provide control independent of the ICS, on loss of ICS power. These are 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 h
Q 1. Implement modifications to allow manual flow control of Auxiliary Feedwater flow independent of ICS in the Control Room.
.2 Actions to be Initiated Prior to Restart and Completed on a Timely Basis
- 1. A detailed schedule will be prepared to addres O accelerated EFIC installation. 't 90%
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3.b Action to Improve Auxiliary St9At kp,@@3 Valve Operation on Loss of ICS Power p g\
Loss of ICS power during the 12/26/85 event opened the auxiliary steam pressure control valve to 507., causing auxiliary steam header overpressure and lifting of one of the header relief valves.
The objective of these actions are to effectively address this plant design condition.
l
.1 Actions to be Completed Prior to Restart
- Assure procedures identify isolation requirements on y) loss of ICS power.
.2 Actions to be Initiated Prior to Restart and Completed on a Timely Basis
- 1. Develop and implement plant modifications to the auxiliary steam controls to assure valve control on loss of ICS power.
3.c Actions to Reduce Main Feedwater Contributions to Reactor Trips b
Like other PHRs, the main feedwater system is a significant 9
g' contributor to plant trips and transients. Theobjecpt .
'v %% g\
5-32
ec '
these actions is to reduce the number and
\ induced trips and transients. gt.
- eedwater
.1 Actions to be Initiated Prior to Restart and Completed on a i Timely Basis
- 1. Perform an operational review of Rancho Seco main feedwater system.
- 2. The recommendations contained in the B&WOG Availability Committee Report, 47-1159449-00, "MFH Pump Trip Reduction Program Final Report" will be evaluated for applicability to Rancho Seco.
- 3. Evaluate potential feedwater systems improvements.
Y p
. ,.t..
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5-33
. - . , _ , . _ , . - - ~ - __ _. _._ , _ _ _ - . _ . - _ _ _ _ _ _ - _ _ _ _
g?
58.1.4 PERFORMANCE IMPROVEMENT AREA ESEL GENERATOR RELIABILITY
'd ,N The reliability of emergency Diesel generators is important to the mitigation of certain events.
The objective of these actions is to enhance the reliability of the Rancho Seco Emergency Olesel Generator systems. To accomplish this objective, the following actions will be taken.
.1 Actions to be Initiated o rior to Restart and Completed on a Timely Basis
- 1. Evaluate the performance history of the Bruce-GM Emergency Diesel Generators and the recommendations for reliability improvement.
- 2. Determine and implement identified modifications.
- 3. Develop a comprehensive preventive maintenance program.
.2 Long-Term Implementation Actions or Actions of a Continuing Nature
- 1. Monitor, trend, and evaluate diesel generator performancg gS, p-O 5-34
_ . . _ . _ , _ _ . . _ . , . _ _ _ _ _ _ _ _ _ . _ _ , . . . . _ _ _ _ . _ . , . _ - _ _ . _ _ - _ . _ _ _ _ _ . _ _ _ . _ _ . _ _ , _ . . . _ , _ . _ _ _ . . _ ~ .
b - 1 Si 58.1.5 PERFORMANCE IMPROVEMENT AREA: PRES CTOR COOLANT SYSTEM (Jn} -
5.a Upgrade Pressurizer Relief Valve Discharge Piping Supports 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 oa July 29, 1983. As committed, these supports will be upgraded to restore their design margin.
.1 Actions to be Initiated Prior to Restart and Completed on a Timely Basis n
- 1. Issue ECNs for new supports and support modifications.
- 2. Inspect, reanalyze, and redesign (as required) ring structure anchoring supports to Pressurizer.
4
- 3. Construct new supports and modify existing supports and ring structure (if required).
- 4. Construct pressurizer support structure q modifications and modifications to existing work gs s @g q60
platforms required to resist new pipe su t@
loads. g 5-35
b d'
58.1.6 LITY O ENHANCETHEPOSTACCIDEN 1 The existing post accident sampling system is difficult to operate and maintain.
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
- 1. Complete the SCAS panel rebuild.
- 2. Complete associated peripheral equipment upgrades.
O 3. Document the compensating equipment in the environmental lab.
- 4. Complete work required to solve H2 monitoring heat tracing problems.
- 5. Revise operating procedures and complete training on revised system and conduct system functional test.
.2 Actions to be Initiated Prior to Restart and Completed on a Timely Basis
- 1. Replace dionex program controller, installation R-150449 cS@
sample dryers. Tb 5-36
N
.! 0 y4 >
- .3 Long-Term Implementation Actions or ACA18n ontinuing Nature %
4 I
4 Complete PASS decay heat valve replacement during the cycle 8 .
outage.
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58.1.7 ACTIONSTOENHANCECONTROLROOM/TSCANDNSEBHVAC-OPERABg(6M 0%
RELIABILITY Sib gg\0 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 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 of the
, systems.
1
! .1 Actions to be Completed Prior to Restart
- 1. Excessive HVAC noise affecting Control Room Operator i
communications l'
- a. Prepare and implement a detailed action plan / testing program to identify source (s) and propose modifications.
I
- b. Develop and implement those design and procedural pd changes needed to reduce noise levels to allow S limits. i gg b g gs\U J $Y 5-38
b Actions to be Initiated Prior to Restart and Complete?
c e'
.2 Timely Basis i gSS
- 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.
- 3. Develop and implement the changes necessary to facilitate maintenance of Control Room /TSC HVAC equipment (i.e.,
replace Air Handler Unit access doors, modify the lube manifold to condenser fans, etc.)
- 4. Fire Damper upgrade actions
l i
- b. Develop and implement the changes necessary to upgrade dampers in the wall between Control Rocm and TSC.
- 5. Control Room /TSC essential filter unit flow control ibW p GU'
- a. Investigate flow control through filter $i recommend improvements, 19, g 5-39 l
[
i 1
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/
ELIMINAONLY -
& PR l 6 E55enty NVAC PURPOSES 2 'C* pre,50r Ti fNhTIVEDISCUSSION FOR 1
j a. Investigate replacement of existing motors with more reliable and higher quality motors (on an as-fall i
i basis).
i
- 7. NSEB essential air handler air flow 1
l '
- a. Develop improved methods to adjust air flow, i
i
- 8. If required, develop and implement design change to add a manifold to each of two units.
I l
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58.1.8 PERFORMANCE IMPROVEMENT AREA
$p AIR SYSTEM RELIABILITY 4 8ch V ,
Rancho Seco has experienced at least two transients in the past due to a 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, including the ADVs.
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
- 1. Complete IAS system review to identify hardware n v modifications required to improve system reliability.
i l 3. Replace leaking letdown filter valve operators.
I
- 4. Add diesel-driven air compressor.
- 5. Provide N or bottled air backup to critical valves.
2
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5-41
- 7. Developandinitiateappropriatemodificationsideg 4 during system review. Q@
4 s
.2 Actions to be Initiated Prior to Restart and Completed on a l Timely Basis 1
- 1. Update P& ids i
l 2. Develop and implement modifications identified in IAS 1
l review.
i, l .3 Long-Term Implementation or Continuing Programs
- 1. Develop and implement modifications identified in IAS l
, review.
l 1
i l
l l
l 5-42
?
58.1.9 PERFORMANCE IMPROVEMENT AREA: HUMAN $D Human Factors reviews of the transients in October and December, 1985 has identified the need to for plant modifications, and procedure revisions.
The objective of these actions is to correct the deficiencies identified and enhance human interface with the plant.
.1 Actions to be Completed Prior to Restart
- 1. Implement modifications and procedure changes resulting from the post trip Human Factors reviews:
4
i
- c. Improve interface between Security and Control Room personnel.
- d. Install long cord on red phone. .
. g.>g{ ,
90 h
5 43
b I . 2 Actions to be Initiated Prior to Restart and Completed on QM Timely Basis \
- 1. Change operator logs to record pump / motor oil level during shift walk throughs.
- 2. Modify Control Room access doors such that one door is used for exit and one for entrance.
- 3. Modify red phone power supply to eliminate spurious ringing.
- 4. Assess capability to accelerate implementation of CROR modifications (Mod 142) currently scheduled for Cycle 8 and cycle a refueling outages.
f I
i . 3 Long-Term Implementation Actions or Actions of a Continuing Nature 4
- 1. Participate in the INPO Human Performance Evaluation program.
ON
- 2. Implement corrective actions identified in December b i %
HED Submittal to NRC. gt,
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J 5-44
- - , . . , - - - .-,-,m_ , . _ , . . , ,. . , _ _ , , . - . . _ _ _ , _ _ - . . _ _ _ - , _ . , _ . . . . . , , _ . . _ - _ . . - _ . - . . - . _ _ , , , . _ . _ _ _ _ _ . ,
, Sb o 58.1.10 PERFORMANCE IMPROVEMENT AREA: E1!L v$es' C 10CFR50APPENDIXIDISCHAg#
,1Y The 10CFR50 Appendix I limits 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.
.1 Actions to be Completed Prior to Restirt k 1. Evaluate the current radioactive waste analysis methods and sensitivity relative to their ability to support 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.
b
- 3. Review and revise the off-site discharge calculation ma t's
.s.-....-.t...s...,.s.
p- .
O 5 45
.2 ActionstobeInitiatedPriortoRestartandCompletAdlOfi A
Timely Basis $f g.
1
- 1. Evaluate the design of plant systems with the intent to improve the ability to operate within Appendix I limits i
- when operating with primary to secondary leakage.
Implement plant improvements as appropriate.
i l
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b 58.1.11 PERFORMANCE IMPROVEMENT AREA: REACTOR BUILDING PURGE F S SAT MEASUREMENTS b The ability to determine containment purge flow rate is currently impacting the off-site dose calculations.
The objective of these actions is to enhance this calculational capability.
.1 Actions to be Initiated Prior to Restart and Completed on a Timely Basis
- 1. Perform engineering evaluation of system deficiencies and identify appropriate modifications.
- 2. Install and test modifications.
O 5-47
_ , _ = _ - -- - - - _
i b6 Y D
- 9 B-2 58.1.12 PERFORMANCE IMPROVEMENT AREA: FIRE ALARM SYSTEM P (ON' Q
The current fire alarm system contains deficiencies in the area of I
operator diagnostic operator cverride and reliability to minimize spurious signals.
i The objective of these actions is to enhance the performance of the fire alarm system and resolve these deficiencies.
Actions to be Initiated Prior to Restart and Completed on a
.1 Timely Basis I 1. Develop and implement the necessary modifications to provide for manual operator override of Auxillary Building
\ ventilation fans.
l
! 2. Upgrade the control logic and schematic diagrams for the fire protection system.
1
- 3. Evaluate the need to upgrade the power supplies to the bi
- existing paneis to prevent spurious signais on power ,e s g.y.erq.,j transients. Develop and implement appropri M 5-48 s
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58.1.13 PERFORMANCE IMPROVEMENT AREA: SEPARATION OF NSEB DAMPER CONT b id '
S CIRCUITS AND EQUIPMENT The current control circuitry and equipment is scheduled to be upgraded to improve its reliability and effectiveness of this equipment.
The objective of these actions is to improve the control circuit and equipment reliability and effectiveness.
i
.1 Actions to be Initiated Prior to Restart and Completed on a l
Timely Basis l
- 1. Develop and implement fire alarm and HVAC panel upgrades to l
separate the control circuitry and equipment for the Train (c. @D '
A and Train B Dampers. 1 g$
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i 5-49
58.1.14 PERFORMANCE IMPROVEMENT AREA: FIRE DETECTION IN NSEB CABLE QS
, AND TUNNELS $\ gg%
The fire detection system in the NSEB Cable Shaf ts and Tunnels needs to be upgraded to improve the reliability to activate the water-spray fire extinguishing systems when required.
I The objective of these actions is to satisfy this need for upgrade to achieve the desired performance level.
.1 Actions to be Initiated Prior to Restart and Completed on a Timely Basis i
- 1. Develop and implement the necessary cross-zoned detection s capability in the NSEB cable shafts and tunnels to achieve the objective of this task.
1 I
4
. O 4 5-50 1
, . , , .. -_ _-.- ,.._. ,.v-. .,.---.r..,__ , -
, g y..,,._____ , _ _ _ _ _
y.._, ...-_ . _ . , , , , _ _ . . _ . - _ _ , . _ , . . , , - . . . . _ . _ . , . _ , _
b 58.1.15 PERFORMANCE IMPROVEMENT AREA: WATER LEAKAGE THROUGH F 9 N s ACTUATION OF FIRE PROTECTION EQUIPMENT A \\
The potential currently exists for water from fire extinguishing systems to leak through floors,and impact safe shutdown equipment.
The objective of these actions is to review and identify areas where safe shutdown equipment could be adversely impacted by water from fire protection system actuation.
.1 Actions to be Initiated Prior to Restart and Completed on a l
Timely Basis
- 1. Identify vital areas of potential impact due to leakage.
1 2. Evaluate effect of impact of potential leakage on safe shutdown equipment.
- 3. Inspect all vital electrical equipment areas for potential l
leakage paths.
- 4. Evaluate the results of the inspection and identify recommended corrective actions.
b 9
d'
- 5. Developandimplementchangesnecessarytoaddresgg recommended corrective actions. gt g\SN 5-51
- 6. Review and upgrade as necessary preventative m.11ntenance procedures to maintain drain lines clear of obstructions, p6.
pBNc @
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5-52
998U %L 5B.1.16 PERFORMANCE IMPROVEMENT AREA: MOTOR OPERATED g$DN O
A near term program is to be implemented to assure all applicable j motor operated valves maintain the environmental qualifications and I
the HPI and AFW systems have valve operator switches selected, set, 1
tested, and maintained properly. Key applicable features of this program are to be extended to other motor operated valves in the
- longer term.
The objective of these actions is to assure reliable operation of motor operated valves is achieved.
.1 Actions to be Completed Prior to Restart I
V 1. Complete applicable portions of the District commitments to
! the NRC in accordance with Mr. John E. Ward's (SMUD) letter l
JEW 86023 of 5/16/86 to J. D. Martin (NRC).
.2 Actions to be Initiated Prior to Restart and Completed on a Timely Basis
- 1. Ccmplete the balance of the District commitments per letter JEW 86023.
.3 Long-term Implementation Actions and Actions of a Continuing Nature
- 1. Extend the motor cperatec valve program to include otheg g motor operated valves.
5-53 g\gg@'g@
b I
58.1.17 PERFORMANCE IMPROVEMENT AREA: CRITICAL PUMPS FAILURE ON LOSS S9 er 93 bi SUCTION y t
j During the December 26, 1985 event, an operator error resulted in 1
}
both suction valves to the make-up pump being closed resulting in 1
) major pump damage.
The objective of these actions is to effectively address this issue to avoid reoccurrence.
i
.1 Actions to be Completed Prior to Restart
- 1. Evaluate procedures and training to prevent recurrence.
I
+
l
.2 Actions to be Initiated Prior to Restart and Completed on a l v
- Timely Basis i 1. Engineering is to review design philosophy for suction valve interlocks on critical pumps and identify appropriat i modifications. qt.
5 4 i
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5-54 l
l l_-- _ _ _ _ _ _ . _ - _ __ _ _ . _ . _ _ _ _ _ . _ _ _ _ . _
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58.2 PERFORMANCE IMPROVEMENT AREA: MAINTENANCE PROGRA g, The quality of the maintenance programs has a direct impact on the sf#*9 e
material condition and reliability of systems and components
- throughout the entire plant. The actions described below are
> intended to provide Olstrict 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 i
I i
- 1. Inventory Calibrated Test Equipment (CTE) and calibrate and/or control use to prevent use of uncalibrated CTE.
l 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).
b
- 5. Perform refueling interval surveillance of snubbers. t O '80'
- 5-55
- . . - _ - - - - . . - . . . . . _ , , - . - . - _ - _ - - _ _ _ - ~ - . . - . _ _ . - _
b
- 6. Complete rework of terminations in the Bailey Cabinets in 6 0 y the Control Room (NNI/SFAS/RPS/ICS). g\ g%
- 7. Perform blannual Diesel Generator Inspection and replace turbo chargers.
- 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 Limotorque 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.
.2 Actions to be Initiated Prior to Restart and Implemented on a Timely Basis
- 1. Develop a departmental procedure hierarchy and writer's guide for Maintenance Procedures.
- 2. Identify and prioritize procedures for generation and/or i revision, qt 969 I g$
5-56
- 3. Achieve authorized staffing levels within t %i g organizations. gS9 l
- 4. Develop and/or revise the required programmatic procedures ,
for the PM program to: assign 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 equipment. Revise and augment as required by programmatic selection criteria. .
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58.3 PERFORMANCE IMPROVEMENT AREA: T g 0$
o 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 Progra:ns up to a state-of-the-industry condition. Since the intent of this type of action is to provide long term rest.lts, there are no items to iI \
(j be completed prior to restart.
- 1. Actions to be Initiated Prior to Restart 1
- a. Continue the upgrade of Non-Licensed Operator Training to achieve INP0 Accreditation.
- b. Initiate the process of achieving INP0 Accreditation i
for the maintenance training areas.
/
i c. Develop plan for the installation of a computeriz .og
$. N Training Information Management System.
l lo s-sa i
k
i
- d. Initiate plans for centralized and secure storage bN
? @Y '
training records. 9 99. E 9g?
1 9. N ef,6 9 10 gg
- e. Develop or purchase a Rancho Seco Simulator Baseline Data Information and Tracking System.
f
- f. Fastor short term action items for specific instruction into long term training material, i
J
- 2. Long Term Implementation Actions and Actions of a 1 Continuing Nature
- a. Complete the purchase and installation of a plant specific simulator.
(
' b. Complete the staffing of the Training Department with SMUD employees.
j
- c. Complete and maintain INP0 accreditation for the l remainder of the Training Programs.
i
.2 Specific Instruction Actions Items - Operations The specific instruction action items for Operations are 93sYbg-
< summarized below: S i gS@D 4 9\
i i
- 5-59
PRELW N' g
pygp0$ES I Actions to be Completed "IOT to Rest
- a. Train licensed operators on Emergency Operating Procedures including the changes resulting from the
! December 26, 1985 event, those revisions resulting from the E0P/AT0G review, and all design modifications that have been incorporated.
- b. Train licensed operators on the loss of ICS/NNI, i including those procedures added or revised as a result of the December-26,1985 event or design modifications.
I c. Train operators on valve applications and operation V including 0]T on local and/or manual operators. This includes limits and precautions such as use of valve wrenches.
- 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 actuator controls.
- e. Train operators on watch standing principles,
[
including command and control training for shift 5-60 #[ 1
. supervisors, role and function of STA, equipme t g fB e t monitoring, etc. 59,,8 ggSSM
- f. Retrain operators on health physics requirements.
.3 Specific Instruction Action Items - Health Physics The specific instruction actions for Health Physics are summarized below:
- 1. Actions to be Competed Prior to Restart
- a. Train Health Physics Technicians on the procedure (s) for entry into areas of unknown radiological i , conditions.
- b. Train Health Physics Technicians on proper response to radiological emergencies.
l c. Train Health Physics Technicians on evaluation of l radiological effluent discharges.
I
.4 Specific Instruction Action Items - Emergency Planning
- p The following specific action items are to improve emergency99$.b GN response. $
1\ g% N l
f 5-61
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1 P66 0 \
f
- 1. Actions to be Comoleted Prior to Resta A 5N E & g gp0SES
, hi$63
- a. Train appropriate maintenance personnel on the
- maintenance of IDADs.
- b. Train operators on the operation of IDADs.
j c. -Update Emergency Preparedness Training Instructor
- i j Guides, Student Guides, and visual aids.
- d. Train appropriate personnel on the revised Emergency Plan Procedures.
l I
i
- e. Provide management guidance (through training) on fire j response actions when plant conditions preclude prompt I
! dispatch of the Emergency Team.
i f 2. Long Term Implementation Actions and Actions of a Continuing Nature l
- a. Develop and implement an improved process for
! continuing Emergency Response Organization Trainin L gg '
and retraining. 1 ggi i
i 5-62
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58.4 PERFORMANCE IMPROVEMENT AREA: OPERATIONS
sgi\NN***# p@i O Improvement in operations is brought about by improving management controls such as procedures, organization, drawings, and through improved training of personnel. Improvements in training are addressed in Section 3 of this Appendix. The actions to improve the management controls are addressed here.
.1 Actions to be Completed Prior to Restart
- 1. Issue management guidani:e, via procedures, defining the policy on procedural compilance. Thl.s procedure will provide a direction on what constitutes " procedural compliance" and " procedural guidance".
- 2. Review the December 26, 1985 transient to identify and correct specific procedural deficiencies.
- 3. Review and upgrade the CR/TSC HVAC operating procedures.
- 4. Verify technical correctness of E0P changes made since May 1985.
- 5. Compare E0Ps to ATOG Technical Basis and incorporate g \N -
significant improvements into EOPs. 9@i 16N t$
O 5-63
MMA8I~
7ggTATWE p RPOSES 0"
- 6. Make the necessary modifications to tha.50%$Ne change
$ process to assure that design changes are incorporated into all operating procedures in a timely manner.
- 7. Assure operating procedures address the recommended topics of Regulatory Guide 1.33 Sections listed below. Implement j procedures which may be required.
1
.1 Section 3 Procedures for Startup, Operation, and i Shutdown of Safety Related PWR System.
J 1
i .2 Section 6 Procedures for Combating Emergencies and i
Other Significant Events.
4
( 8. Verify configuration of important secondary mechanical i systems, perform a valve walkdown to verify the consistency I
of as-built conditions, P&I0s, procedural lineups, and component identification. Initiate corrective action for i,
inconsistencies which are identified. This compliments the 1
long term Configuration Management actions defined in 1
Section 10 of this Appendix. Systems included are:
l Air Ejector / Gland Seal j
/
M Auxillary Steam sess ggggg ?@i& cd '
i Component Cooling wat gg5 instrument Air s\W 5-64 l
t
0.' \
b PB6
\M6 $050 Main Circulating WateTr6 gSCOSS\ ,
- Plant Cooling Water i
Service Water Turbine Electro Hydraulic Control Turbine Lube Oil i
.2 Actions to be Initiated Prior to Restart and Completion on a
)
- Timely Basis
- 1. Approve a revised Nuclear Operations organization and begin ;
i staffing at.the management level.
l 2. Develop a staffing plan and schedule to meet the needs of I
- the revised Nuclear Operations organization. This will
,666i
! includetheneedsforlicensedoperators1{egjgt g qd "
rotationau transfer assignmen g g o $ *ss e ? #
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&9 QS gg, N 58.5 PERFORMANCE IMPROVEMENT AREA: EMERGENCY PREPARED BSS
- (
i
! 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 i
j weaknesses were apparent in the Emergency Preparedness Plan, and its i
l Implementation during the December 26, 1985 event.
l 4
The objective of these actions is to upgrade the Emergency
- Preparedness program to assure that it is efficient and effective.
l
- .1 Actions to be Completed Prior to Restart l
I,
! 1. Develop and implement the necessary changes to incorporate the lessons learned from the December 26, 1985 event. !
f i
i 2. Review and evaluate the on-site and off-site Emergency
) Preparedness capabilities to determine if additional hardware or support is required to achieve the objectives i
of the program.
i
- 3. Review and evaluate the Event Classification Guide, and the l
! Emergency Response Procedures, to determine the need for changes to achieve the program objectives.
i
- 4. Review and evaluate the communications and notification g '
l perrormance during the oecemeer 26, 1985 event, and 3 '"*
l gg$e*e#gi$
5-66 ,
i
, - - - - , , , , , - , _ , . - ~ . - ~ . - _ _ _ - _ - _ _ . - , . _ , _ _ . --- __m_ _ - . _ , _ _ , -
E g tiss10M P R previous event drills, or responses. Determine the O corrective actions necessary to achieve performance consistent with the program requirements and good practices.
- 5. Conduct the necessary personnel training to assure proper implementation of the Emergency Preparedness program.
- 6. Conduct an on-site drill, following the implementation of the above program improvements, to confirm their effectiveness.
.2 Actions to be Initiated Prior to Restart and Completed on a Timely Basis
- 1. Develop and implement actions to achieve and maintain a high level of performance in the Emergency Response Area consistent with the INPO performance objectives.
- 2. Develop and implement the design changes and procurement activities necessary to obtain the hardware or achieve a hardware configuration supportive of the program. \N I 9 D ggg &
15$sesS*' post.S i
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5-67 l
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PBN L' 58.6 PERFORMANCE IMPROVEMENT AREA: QUALITY AND QUALITY A g (? N g
v An effective quality program at Rancho Seco is important to achieve the near and long-term performance standards of the District.
The objective of these actions are to improve the overall effectiveness of the Quality Assurance area and to assure the benefits of quality assurance are reall:ed in the near and long-term.
.1 Actions to be Completed Prior to Restart
- 1. Develop and Implement the procedures and processes necessary to independently verify the effective closure of the actions identified as requiring closure prior to t
(,/ restart.
.2 Actions to be Initiated Prior to Restart and Completed on a Timely Basis
- 1. Update and modify the Quality program policies and procedures, particularly those dealing with material control, engineering, quality surveillance, and maintenance.
l
- 2. Identify major programmatic and management areas that g@
S '# $
require enhancement 1 from a quality o
o 5-6a
i i
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.3 Long-Term Implementation Actions and Acticns of a Contigg & 98 gsGud '
ew Nature 99g Olsgss e e *?g i
- 1. Develop and implement the necessary policies and procedures to establish a more proactive quality program and improve the effectiveness of audits.
i i
5-69
PBN c \. '
58.7 PERFORMANCE IMPROVEMENT AREA: MANAGEMENT EFFECTI & ggg980 Management and management process effectiveness have a major impact on the ability to operate the Rancho Seco Nuclear Generating Station
'in a safe and reliable manner.
The objectives of these actions are as follows:
- a. To enhance the management process in support of the safe, reliable and timely restart of the Rancho Seco Nuclear Generating Station,
- b. To develop guidelines and agreements by which the SMUD Board, as a governing entity, can improve its effectiveness in directing O and monitoring the District's activities relating to the Rancho Seco Nuclear Generating Station.
- .1 Actions to be Completed Prior to Restart i
i l 1. Review and ensure that executive level nanagement processes support the safe, reliable, and timely restart of the Rancho Seco Nuclear Generating Station, i
.2 Actions to be Initiated Prior to Restart and Completed on a l
Timely Basis f
I , geww%f-g i. ecare 0,,ec,ers, Gene,al Mana,e, ,m,,, g ,, ,
- 5-70 1
Y TENTATIVE Estab;;gy WIthin the go"'d guldenjy DISCUS $
OSES CNLY PR
- and ag'***ents 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
- a. Develop and implement a program to provide the Board and General Manager assessments and recommendations of
! corporate management improvement.
l l 3. Nuclear Program Management I
a, Develop and implement a Rancho Seco Business Plan for use by the Board of Ofrectors.
i b. Establish a comprehensive, cohesive and clearly understandable set of GM and AGM-Nuclear policies and practices which provide upper tier direction for g(p similar efforts at the functional manage n1p , C'00 '
su e,v,,e,, levei,.
g ,se 5-71 L
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1 6 b'
- c. Establish up-to-date functional organI$ tion charters se n s #PB
!k and position descriptions which accurately reflect responsibilltles authorities, and accountabilities for all organization functions and jo' u classifications.
4 l
1
- d. Upgrade management programs and practices in the areas of functional planning, decision making, problem I
j solving and interdepartmental collaboration.
i i e. Establish appropriate management monitoring and l
l control systems to ensure that all levels of department management are kept informed on important I department performance trends or problem areas on a
- g.
timely basis. At the same time, ensure that excessively burdensome administrative control systems
- are not perpetuated or introduced.
I
! f. Develop an employee communications program originating i
! from the office of the AGM-Nuclear to ensure that all
! department employees are kept informed of Olstrict 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, i
i
- g. Develop a program for improving communications skills j 8 I of Nuclear Department managers in presentations g4 -
Board of Ofrectors, the public, and sta g46 967 l
$$5#
i 5-72 1
I
i pBE \, i T@ \@
- h. Establish a department Human M rce Management iA
- program which includes:
i j 1) Identification of priority management i
i development / training needs and the appropriate means for addressing each; and l
- 2) Identification of departmental priority in terms l of current vacancies and/or pipeline concerns and 1
i more department management collaboration with the Olstrict's Human Resources organization in the recruitment / selection process.
I l
- 1. Improve Department media and community relations by
)'
establishing a more proactive media / community outreach j program.
l i j. Improve Nuclear Department interfaces with all other Departments in the Olstrict by instituting additional
! Interdepartmental communication and problem-solving
- processes on a regular basis.
l
- k. Facilities improvement needs are to be establish i, i
sg>
I i
5-73
g?
58.8 PERFORMANCE IMPROVEMENT AREA: CCMMITHENT MANA 90% i Commitment management is an important aspect of the 01 strict's interface with "outside" agencies as well as for the management of the Olstrict's day-to-day activities.
Impravement in commitment management is aided by improvements in management controls, such as procedures and tracking systems. The actions to improve the commitment management controls are identified below.
.1 Actions to be Completed Prior to Restart
- 1. Review and revise commitment management procedure.
! d
- 2. Install new commitment tracking system.
J 4
- 3. Develop system / user documentation for the new commitment l
tracking system.
I h
l 4. Verify the commitment tracking system database with respect i
i to current commitments.
l l ^
.2 Actions to be Initiated Prior to Restart and Completed on a j
Timely Basis
- 1. Verify the commitment tracking system histcrical commitmgg\@
database. gg g%@?
1 f 5-74
pBEU I s\t Y g-TEN \% N
- 2. Each Nuclear Department will establish specific milestones for reducing its backlog of open commitments.
.3 Long-Term Implementation Actions or Actions of a Continuing Nature
- 1. Integrate the commitment tracking system with the Nuclear Information and Data Management System.
1 i
1 l
l l
5-75
PB60 \.
58.9 PERFORMANCE IMPROVEMENT AREA: RECORDS AND DATABAS 4 Op 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 efficiently and effectively support the Nuclear Department's needs.
.1 Actions to be Completed Prior to Restart
- 1. Review SMUD programs to address vendor documents and recommend enhancements.
- 2. Provide site information systems support to support the implementation and records management activities for restart.
3.
Prepare a Data Systems Olrectory of existing systems w tgggM ,
data source information. p\4 y@?D 10 5-76
PBEU ' L
.2 Actions to be Initiated Prior to Restart and ..
I hna Timely Basis
- 1. Evaluate Vendor Data Program, develop and implement actions to achieve the program objectives.
- 2. Consolidate the stations for control documents.
.3 Long-Term Implementation Actions and Actions of a Continuing Nature
- 1. . Establish program and policy for development of integrated data management system design and implementation.
I
- a. Complete NDMIS evaluation
- b. Establish on-site facilities and organization to support NOMIS hardware / software l kB
\ P @d '
l C. Ianplement NOMIS program / data 3gqq. gp
! 19.N ef,6N
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E? ot 58.10 PERFORMANCE IMPROVEMENT AREA: CONFIGURATION MANA hi g@ i
% i Configuration management identifies, controls, statuses and verifies 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.
The objective of these actions is to take the necessary steps to improve the effectiveness of the configuration management system at Rancho Seco.
A
.1 Actions to be Completed Prior to Restart
- l. Verify that control room drawings are updated in accordance with existing procedures. ,
- 2. Provide Nuclear Engineering support to plant operations as ,
required to address configuration management issues.
.2 Actions to be initiated prior to restart and imolemented on a ,
s timely basis s
i
- 1. Review and evaluate all temporary modifications and close ,_
m out all existing abnormal tags that need to be convertgg\M permanent plant modifications. qq,$T g$ ?N 5-78
,=.m
I gi%\NN ?
g 2. Reduce the backlog of DCNs to level consijtGB th ater).
l
.3 Long Term Implementation Actions and Actions of a Continuing Nature
- 1. Establish management direction for a Configuration Management program for Rancho Seco consisting of:
- a. Policy
- b. Specifications
- c. Computer hardware and software
- d. Implementing procedures
- e. Training
- 2. Establish or upgrade existing equipment and supporting documentation identification systems needed for total plant configuration control.
- 3. Develop total change control packages that control modification from change request through close out. These includes identification of all affected documentation such as procedures, training plans, and simulator upgrades.
- 4. Reorganize drafting into a design / drafting organization.
- 5. \.
TrainNuclearEngineerstoutilizethesedesignersgap~B6
"" '" " "" " "' " ' '"~ ~*"##'*"
O l 5-79
S gpWF- g p ?CS
- 6. Develop new, or simplify existing pto669u9 o clearly define the review process for drawings.
- 7. Conduct a cost / schedule review to determine whether a CAD system can be justified for Rancho Seco.
- 8. Develop or upgrade existing systems to provide verification 4
that configuration documentation reflects the true hardware configuration.
- 9. Develop or upgrado 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 cha gg.ppy.ON -
gs ecc s
i 5-80 l
PBf- 0 \
- 58.11 PERFORMANCE IMPROVEMENT AREA
- HEALTHPHYSICSANDR)
O -
V CONTROLS Improvements in Health Physics or radiological controls is brought about by improvements in organization, procedural controls, and training. Training is addressed in Section 3 of this Appendix.
Organizational and procedural actions include:
.1 Actions to be Completed Prior to Restart
- 1. Relieve operators of special HP duties.
- 2. Prepare a procedure, for Health Physics Technicians to use for entry into unknown radiological conditions.
l 3. Revise setpoints for plant gaseous effluent monitors to ensure unambiguous indications.
l 4. Issue a Radiological Event Directions Manual to provide more guidance for abnormal situations. Issue new manuals ,gg to separate event and instrument procedures frg gif g S QUd '
Radiation Control Manual. I g gSCDSS i
O 5-81 b
inI 5B.12 PERFORMANCE IMPROVEMENT AREA: MATERIAL MANAE $
C The Material 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 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 Prior to Restart and Completed on a Timely Basis
- 1. Conduct a review of the current Materials Management program.
- 2. Develop and implement an action plant to improve the gphN gd '
performance of the Materials Management progr qq, & p 99 1*fdeSGS' O
5-82
/N 6.0 O O TEST PROG %
- 6 tt9h District is developing a restart test program, to augment the other activities identified in this document, which will assure District management that Rancho Seco can be safely and reliably returned to service. The guiding philosophy for the test program is that, when it is completed, it will have demonstrated that important systems will perform in a manner consistent with their design intent. In particular, the program is designed to assure the operability of those systems required by the Rancho Seco Technical Specifications as well as demonstrate acceptable material condition and functionality of those systems (not covered by Technical Specifications) which can contribute significantly to reactor trips and challenges to safety systems.
t U
O)
As a means to ensure the District has optimized its ability to learn from other programs, personnel from the Rancho Seco test organization will visit Davis-Besse to review their test program. The intent of this visit is to identify enhancements, problems, and failures encountered by Toledo Edison in their restart test program. These lessons learned will be evaluated for applicability to the Rancho Seco test program.
In addition, as part of the refinement process for the Rancho Seco Test Program, the Davis-Besse Test Program will be reviewed with regard to its general objectives and methodology. This review, coupled with the evaluation of the lessons learned through the Davis-Besse Testing /
Program will provide additional assurance to the District that its O
/ program's objectives and methodology will provide a souno basi
('
resuming normal plant operation.
6-1
b ON I 9-19 Werall test program is a combination of the surveillance test G M- N program, which demonstrates the OPERABILITY of those systems covered by the Rancho Seco Technical Specifications, component testing associated with the implementation of Design Change Packages, routine post maintenance testing, and Special Test Procedures. Special Test Procedures are procedures prepared especially for certain major maintenance activities, modifications, or specific performance testing for this Restart program.
Test Objectives and Acceptance Criteria will be included in each test to assure that the test, or combination of tests, demonstrates the specific system operability or functionality, as intended.
An additional objective of the test program is to ensure that the surveillance program is current and will support power operation until the next scheduled refueling outage, Cycle 8.
As a result of this objective, the following major surveillances will be completed as described below. These significant activities provide additional assurance of acceptable material condition throughout the facility.
A full' pressure (52 nsig) short duration ILRT will be performed.
This will be the fou th full pressure test. Previous tests were performed prior to initial startup, and again in 1977 and 1983g9 L TLRT program will also be kept current. \ 4\D m
6-2
O 73 4 e inservice inspection examinations remaining in the initial b O year ISI program, except those requiring removal of the reactor
,5@ vessel head, will be performed. These excepted examinations include portions of RV studs and stud hole ligaments, weld joints in one CRDM, RV head weld, and visuals of valve internals in the DHR system that require defueling. Augmented examinations will be performed that include an RC Pump Motor flywheel, welds in high energy pipe systems and radiography of the original HPI thermal sleeves.
An Integrated Engineered Safeguards Actuation Test.
The Emergency Diesel Generator biannual inspection will be completed.
A by-product cf the test program is the opportunity to verify and/or
\ validate procedures which were modified as a result of design changes.
Table 6-1 is a tabulation of the testing to be performed. This tabulation is segregated by General Objective and identifies, by system, the Scope, Specific Test Objectives, and Test Method to be used.
NOTE: Specific Test Objectives have not been totally defined as of this submittal. It is expected that these will be completed by the July update I
of the plan.
The Test Director will be responsible for determination of the exact test '
method to ensure the Specific Test Objectives are satisfied. For those y b
V cases where Special Test Procedures are required, the gl/
6-3
l l
l
/
dev reviewed and approved in accordance with the facility Technical (
i I
O b fications and management control systems.
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RANCHO SECO RESTART g g tfif.f03f.3 W '
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.
(T0 BE COMPLETED IN FINAL SUBMITTAL)
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Nuclear fuel Core performance Verify core design parameters and develop updated operating curves Routine physics and data. test program.
Letdown and Letdown filter Purification pneumatic valve acceptance Containment Valve performance Building Spray Auxiliary ICS Control feedwater Valve -
operability Reactor System function Protection Safety System function Features Actuation Non-Nuclear System function Instrumenta-tion 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.
Annunciator verification To verify that the termination inspection and rework did not result Special Test in improper terminations by verifying proper relationships between Procedure INPUTS, PROCESS PARAMETERS and OUTPUTS.
To verify that the design change on the interface between the ICS Special Test sources and the Annunciator System were installed properly and the Procedure design intent was achieved.
m __ - _ _ _ _ . _ _ _ _ . _M
m O TABLE 6+ phntinued)
RANCHO SECO RESTAR E & PRELIMINARY PAGE V )2 10N PURPOSES ONLY -
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.
(TO BE COMPLETED IN FINAL SUBMlfTAL)
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE /METHOO
! Integrated To verify proper operation of power supplies. Special Test Control System Procedure l (continued) To verify control board indications on loss of ICS DC power. Special Test Procedure To ensure proper tuning of the system. Normal calibration DC Power Battery capacity Control Room System function Essential Noise level HVAC Cooling capacity System integrity Instrument Diesel driven Air air compressor acceptance l
System integrity Back-up gas I bottles fire Auxiliary Protection feedwater pump Hater deluge system acceptance i
Nitrogen Gas Test nitrogen gas - instrument air crosstle PASS System function i
- __---m _ _ . _ _ _ _ _ . _ _ _ _
( ) TABLE 6 ntinued) ( j PAGE ' '1 RANCHO SECO REMBIIy;3yIL@ PREUMINARY USS10N PURPOSES OHLY -
MAJOR TEST OBJECTIVES: Verification of operability of safety syfth DIan functionality of non-safety systems following maintenance activities.
(TO BE COMPLETED IN FINAL SUBMITTAL)
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Reactor System integrity Verify by Operational Leak Check and system walk-down, that the RCS Surveillance Coolant System Pressure Boundary is intact as required by Technical Specifications. Procedure Seal Make-up pump Verify proper operating characteristics of the Make-up pump Special Test Injection and performance following rework. Procedure Make-up Lontainment Penetration leak Building rate Steam Integrity Generators Main Steam Relief valve setpoint Safety System function features Actuation Integrated Continuing checks Control Module checks functional check Si and S2 trip times Non-Nuclear System function Instrumenta-tion Control Room System function Essential HVAC HVAC Charcoal performance "B" Auxiliary ButIding_ exhaust
TABLE 6 *ntinued)
PAGE 2 RELIMINARY RANCHOSECORTEf5h%INb RPOSES OHtX - .;
FOR B W "
MAJOR TEST OBJECTIVES: Verification of operability of safety systems and functionality of non-safety systems following .
maintenance activities.
(TO BE COMPLETED IN FINAL SUBMITTAL)
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD HVAC Charcoal performance "B" '
Aux 111ary Building exhaust unit t
i i
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L
t TABLE 6 ontinued) l .
[
\~- gpuli PAGE \ 'l RANCHO SECO RESTART TESTING p(\EU gg -
MAJOR TEST OBJECTIVES: Verification of system operability per te3 991ca on.
(10 BE COMPLETED IN FINAL SUBMITTAL)
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Core fload Level indication check valve condition Seal Injection HPI pump and Have-up performance; HPI (HPI) valve performance Decay Heat Pump performance; System valve performance system integrity Containment Pump performance Spray Valve performance Sodium hydroxide concentration System integrity Nuclear Pump performance Service Cooling Water Nuclear Pump performance Service Raw Hater Borated Water CBAST concentra-tion Pump performance Containment Integrity Building (penetrations)
Spent fuel Pump performance Cooling
TABLE 6 ontinued) I (N --
) PAGE i -- 2 RANCHO SECO RESTART TESTING g gg1ARY MAJOR TEST OBJECTIVES: Verification of system operability per TecK6Is \ t( $$
(TO BE COMPLETED IN FINAL SUBMITTAL)
SYSTD1 SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Steam Integrity Generators Themodynamic/
heat transfer performance Main Steam Valve performance Auxiliary Pump performance feedwater Valve performance Flow path flow instrumenta-tion Reactor Input device Protection calibration System calibration Safety System features calibration Actuation Integrated test Integrated Start-up tuning Control program Nuclear Calibration Instiumenta-tion Non-Nuclear Calibration Instrumenta-tion Radiation Calibration Monitoring PASS System function
i TABLE 6 antinued)
PAGE 3 RANCHO SECO RESTART TESTING git 4AM MAJOR TEST OBJECTIVES: Verification of system operability per Te g cf ggaX -
f5l 1A*"' (TO BE COMPLETED IN FINAL SUBMITTAL)
TEST TYPE /METN00 SYSTEM SCOPE SPECIFIE' TEST OBJECTIVES Hydrogen System function i Honitoring t
Selsnic Calibration Monitoring ,
. Emergency Diesel
- Diesel performance ,
Generator Control Room Charcoal I Essential condition; filter HVAC condition; system '
Integrity; system function !
t Normal llVAC Filter testing on selected units j functional test i
on selected units fire Pump performance
, Protection System integrity ,
Hater Detection system '
lI function Carbon Dioxide Control system function Flow path l
r i
I ) TABLE 6 ntinued) .
4 x_ -
PAGE k -
'l RANCHO SECO RESTART TESTING ggit4ABI
& PR MAJOR TEST OBJECTIVES: Verifying acceptable material condition y t %W %Pm gq those Ot{L1 -not covered by Tables IA systems
-ICandwhichcancontributesignifican]t,lpyg or trips, system or operator challenges.
(TO BE COMPLETED IN FINAL SUBMITTAL)
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Nuclear fuel Integrity Core Core performance vs. prediction Reactor Seal condition Coolant Pumps Vibration High Pressure Trip devices Turbine Vibration Valve operability Steam Integrity Generator Thermodynamic /
heat transfer performance Main Condenser Integrity Feedwater Level control Heaters Integcated ICS tuning at Control 101, 401, 751 and 901 power
- TABLE 6 ntinued) !
( ~ PAGE ' 'l RANCHO SECO RESTART TESTIN gittb MAJOR TEST OBJECTIVES: Toensureoptimumsystemperformanceandreg f gMg ongoing testing.
~ ~ ~ ~
3; N (TO BE COMPLETED IN FINAL SUBMITTAL)
,___SYS_ TEM _ SCOPE SPECIM C TEST OBJECTIVES TEST TYPE / METHOD Reactor Flow Coolant Pumps Motor performance Letdown and Valve performance Purification demin/ filter performance Pressurizer Valve / pump Relief Tant performance Nuclear Heat exchanger Service Raw performance Wate Waste Gas Integrity Spent Fuel Heat exchanger Cooling performance filter /demin performance RCS Sampling Valve performance Heat exchanger performance Radwaste None High Pressure Thermodynamic /
Turbine / heat transfer Extraction performance Steam Low Pressure Thermodynamic /
Turbine / heat transfer Moisture performance Separator /
Reheaters
7s
/
! ) TABLE 6 ntinued)
'w j PAGE ' /2 RANCHO SECO RESTART TESTING g%
MAJOR TEST OBJECTIVES: To ensure optimum system performance and reliab I? ng testing.
~
W ' g\$ W ~ (TO BE COMPLETED IN FINAL SUBMITTAL)
SCOPE SPECIFIC OBJECTIVES TEST TYPE /METH00 SYSTEM _
Main Condenser Heat transfer feedwater Thermodynamic /
Heaters heat transfer performance Heater drain pump performance Gland Steam Valve performance feedwater Pump performance Pumps Turbine performance Control performance Main Pump performance Condensate Valve performance and Mole-tip Demineralization performance Circulating Pump performance Water Cooling tower performance Valve performance Canal Station Pump performance Valve performance Traveling screen performance Turbine Lube Pump performance Oil filter performance Centrifuge performance Valve performance
TABLE 6 ntinued)
PAGE ' 3 R_ANCHO SECO RESTART TESTING 94pBi MAJOR TEST OBJECTIVES: Toensureoptimumsystemperformanceandreliabil}%th gs@oNgtesting.
-rftID suct\@
g DW"~ (TO BE COMPLETED IN FINAL SUBMITTAL)
SYSTEM SCOPE SPECIFIC TEST @ OBJECTIVES TEST TYPE / METHOD Air Ejector Ejector performance Condenser performance Gerierator Seal Control valve Oil performance Pump performance Au>iliary Boller Steam performance Control valve performance Turbine None Sarnpl i ng /
Chemical Addition Control Rod Rod drops Drive functional check-out Stator checks Main Generator Performance test
/ Exciter EHC/ Auto Stop Pump performance Oil Control valve performance Filter performance Integrity Integrated Power supply Control load test
~
( l TABLE 6- ntinued) i
'-- 3\Ogi - PAGE
-4 RANCHO SECO RESTART TESTINGg% @
MAJOR TEST GBJECTIVES: To ensure optimum system performance and e 1 W ou h ongoing testing.
@*' (TO BE COMPLETED IN FINAL SUBMITTAL)
SYSTEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Comunica t ion None 120VAC None 480VAC Transformer performance 4160VAC Transformer performance Under/over voltage calibration 6900VAC Transformer performance 12.4LVAC/ None 22LVAC Plant None Computers Normal HVAC Filter testing on selected units Charcoal testing on selected units functional test on selected units Service Air Compressor performance Aftercooler heat transfer Compressor control functional test Air consumption (system integrity)
. ^s TABLE 6- atinued) )
b PAGE V5 RANCHO SECO RESTART TESTING g};N '
b HAJOR TEST OBJECTIVES: To ensure optimum system performance and reliabil oing testing.
@ '"' (TO BE COMPLETED IN FINAL SUBMITTAL)
SYSIEM SCOPE SPECIFIC TEST OBJECTIVES TEST TYPE / METHOD Instrument Air Air dryer performance Filter performance Air consumption (system integrity)
Component Pump performance Cooling Water Heat exchanger performance Surge tank function
~
Plant Cooling Pump performance Hater Heat exchanger performance CRD Cooling System integrity Hater Pump performance Hydrogen Gas None fuel Handling None Demineralized None Water Domestic Water None Service Hater None l
l l
s 1 TABLE 6 V .ntinu:d) PAGE 6 ,
RANCHO SECO RESTART TESTING gj$I MAJOR TEST OBJECTIVES: To ensure optimum system performance and reliability t ogg9 Df' testing.
Pkcp3FBE COMPLETED IN FINAL SUBMITTAL)
SYSTEM SCOPE SPECIFIC TEST MvES TEST TYPE /METHOO Aucillary Pump performance l i
Diesel fuel J
I Drains, Sumps None Demineralizer Pump performance Reactor Demineralization 4 Colant performance Crane and None
- Holsts ,
i l Sewage None Treatment Site Reservoir None Building and None
- Structure ,
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0 ~
7.0 Action Plan Schedule
\ h 9af5f The summary s le for the major actions, which must be complete prior to restart, is provided in figure 7-1. This schedule was developed based on a first cut evaluation of the resources and durations required to accomplish the specific tasks. Additional detail is being developed for the final issue of the document which will improve the confidence in the schedule.
j A key source of uncertainty associated with the schedule is the ;
) generation of recommendations through the Plant Performance and O Improvement program. This program has the potential for identifying actions which have not been accounted for in the plan. Additionally, certain major test program activities, such as the ILRT and other in-service inspections activities, can have a significant impact dependent on the findings of the inspections and tests.
i i .
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\
' 7-1
r L BANCHO SEC01. 4 TART SCHEDULE S
ac- ou m no a. s 20 f
GENIItAH ot D LACIION PLAN l l l yh l l l l l l PREPARE DRAFT l l 5-28l . l l l l l ,
i MANAGEMENT REVIEW AND APPROVAL l 5-2sl 5-29l k l l l l l l FINALIZE DRAFI l 5-28l 5-30l l l l l l l l TRASISMIT DRAFT TO NRC l l l 6-2 l l l l l l l
} ORAFT PRESENTATIONS l 6-2 l l 6-10l l l l l l l REFI%E DRAFT AND SCHEDULE l 6-2 l l 6-27l l l l l l l MANAGEMENT REVIEW AND APPROVAL l 6-27l l 6-24l l l l l l l ,
I fit:ALIZE ACTION PLAN l 6-27l l 6-30l l l l l. l l ,
BOARD OF DIRECTORS APPROVAL l 7-1 l l l 7-3 l l l l l l [
TRA3 SMIT ACTION PLAN TO NGC l l l l 7-3 l l l l l l NPC REVIEW AND APPROVAL l 7-3 l l l 7-18l l l l l l MONTHLV PLAN / SCHEDULE UPDATE l 7-3 l l l l l l l l l l-3 l l l l l l l l l l l
- l l euwLftata Mcm mPROV PBDGRAM l l l l l l l l WEVIEW PROCESS TO GENERATE RECOM. l l l 6-18l l l l l l l .
- REviEwDISPOSITION REC S er RRRs l l l l 7-4 l l l l l l
} PAG APPROVE RECore1ENDATIONS l 5-28l l l 7-15l l l l l l l INCORPORATE INTO RESTART SCHEDULE l 5-28l l l l s-1 l l l l l l l l l l l l l l l l
j Ioranatoannuccincanons l l l l l l l l l l l DESIGN AND PROCUREMENT l l l l l l 9-30l l l l 13:STALLATION l l l l l l l 10-30l l l f 11-30l l j FUNCTIONAL lESTING l l l l l l l l i
l l l l l l l l l l j MAINIEMMCLfkoGRAM l l l l l l l l l l I
j INSTRUMENIATION CALIBRATION PRGH l 6-1 l l l 7-15l 8-1 l l l l l '
I PM PROGRAM CRITICAL EQUIPMENT l 6-1 l l l l s-1 l l l l l
! IEN 86-03 M0v-to- l l l l l l 9-i l l l l IES 85-03 'H0V SWITCH SETTING' l l l l l l 9-20l l l l l
MISC MAINTENANCE ACTIVITIES l l l l l l l l l l ;
- l l l l l l l l l l l omgaucus_mo_esoctoueLIneson l l l l l l l l l l j UPGRADE OPERATING PROCEDURES l 6-1 l l l l. 4-15l l l l l PE4 FORM SELECTED SrSTEM WALKDOWNS l 6-1 l l l l 8-1 l l l l l ,
l l 9-30l l l l
! REVISE PROCEDURES FOR MODS l l l l l l l Data l l l l l l l l l l oate l l l l l l l l
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. .- -. . - - - - . . -- .~ _ - .
- m tLIX An hl aus \N a, oC1 Kor Das Jan IRAINING istE_LPJoR E.P. TRgG) l l l l l l l 1 l UPGRADE TRAINING FACILITIES I i l l 7-1 g h l I I l CONDUCT IMPk0VfD RAD TRAINING l l l 1 k -1 1 I I l l PREPARE FOR MOD TRAINING l l 1 N I 9-30l l l l CotsoucT Moo TRAINING l l l 1,k@ l l l 10-301 1 I I I I I I I I I I I tuikctact_estementss l I I I I l l l l l UPcRAoE EnERcENCv eRoCEouRES I 6-i I i I 7-iSI l l l l 1 TRAIN EP PERSONkEL ON NEW PRCDRS I I I I l 8-271 1 I l l HOLD PRACTICE EHERGENCY DRILL i i I l l 8-281 I I I I cress REnEARSAL FOR FORnAL ExC I 1 l l l 1 9-sol l I I AmeuAt EnERGENcv EXERCISE l l l 1 l l l io-6 l l l l 1 1 I I I I I I l ou w1Ir i I I I I I I I I I IzoEPENDENTVERIFbFRESTRTISSUESl 1 l l 1 1 I i 11-30l l 1 1 I I I I I I I I nAnctntNLIssucs 1 I I i l i l I I i AS;ESSnEuT or ncar oIRECTIoM ANo I l l 1 l l l l l l CocITMENT rc6 RESTART l 6-i l l 6-151 l l l l l l l l l l l l l 1 1 I co! m IntstctnTeot i I I I I I I I l l RE/ISE COnHITHENT CONTROL PROCDRE I 6-i l i l I 8-151 I I I l DEVELOP CON EONTROL DOCUMENTATION l 6-1 l l l l 8-151 l l 1 I VERIFY COtttITHENT DATA BASE I 6-1 l l l l 8-151 l l 1 1 I l l I I I I i ! I accoRosJListoEUATION HANAGtHEMI l 1 l l l 1 l l l l coxGoING> l l l l l l l l 1 1 I I I I I I I I l l CONflGuRAIIOR.HANAGitGI i l l l l 1 l l l l AUDIT / UPDATE CONTROL ROOH DRAWINGS I 6-1 l l l l 1 9-1 I l l l ESTAetISH NuC EN. SuPPoRr group l l l l 7-1 I I l l l 1 1 I ! I I I I I I I nunas facIces_E8GIntisInG l i I I I I I I I I REvlEu nEo S rRon 12/26 EVENT l l l l l 8-i l l l l l 1 I I I I l I I l l ctuteIc_Issuts I l i I l I i l i I (TO BE DElEmlINED) i I I I l l l 1 l 1 1 I nata l l I I I I I I I I cete i I I I I I I I 7-2
- - . - . ~ -. -. -. _- - - _ . _ _ _ - . _ . - - - - - _ - . . . . - . - . - - . - . - . . - . _ . . ~ .. . _ . _ .
i l
Max Juo Jul aus sen Oct. mar ces Jan l
! MinIn entsics i I l l l l g:0f l l l
- Ret!tvt ces or sPECIAL RP DUTIts l l l 6-i l l (\N l- l l l 4
Rev streo w s roR cAstous trr Hon l l 1 6-i l l g gs l l l l t
RE V HAN To IMP cUIDNCE To RP TECns l l l 6-i i l l l l 5@hggf,Ngl
! I l l l l l l l II; L n oGRAM i l l l - l l l l l l l l ii-sol l RESTART TtsT!uc roR Hoos l l l l l l RESTART TrsTING tor HAINTENANCE l l l l l l 9-30l l l l RisTART TESTING roR oP (sP'st l l l l l. l l l l 12-24l eowtR Ascrusion itsTInc 112-241 l l l l l l ! I i-zz o c eRt-ntArue RitcAst l l l l l l l l l iz-i i iz-i l i rinat tIntues lii-i l l 1 l l l l l cso TcsTInc liz-i l l l l l l l l iz-a l
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- RX POWER 10 4u% TESTING li2-3
- l l l l l l l l 1 i-7 RX PoutR To 75% TrsTINC l i-7 l l l l l l l l l i-i4 j RX POWER To 90% TESTING l i-i4l l l l l l l l l i-2:
l RX PoWtR AT 100% l i-2il l 1 l l l l l l i-22 l l Data l l l l l l l l l l t l l oate l l l l l l I
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l NG TERM TRANSITION T.t:NTATIVE & PRELIMINARYURP FOR DISCUSSION P i
j The transition of the special NRC and independent oversite activities i
for this Restart and Performance Improvement Program to those consistent 4 with normal industry practices should occur when the restart actions are complete and those actions to be implemented prior to restart and i
implemented in a timely manner have been initiated and developed in 1
sufficient depth to assure their completion on the projected schedule.
The transition of the special processes such as the precursor review, the B&W Owners Group issues assessment and resolution, and Reliability / Failure Consequence analysis to the line organization will be accomplished following the plant restart.
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. 8-1
b @D 9.0 SPECIFIC DISTRICT RESPONSES TO NU Q6 DINGS V
$t, g\
ThefollowingaretheDistricYsresponsestotheconclusionsand, recommendations section of NUREG-1195. s
.s ,
FINDING - SECTION 9.1 .
- 1. The December 26, 1985 overcooling transient was initiated by the power supply monitor in the nonsafety-related ICS (trippfog the
+/- 24 Vdc power). The most probable catise of the tripping was a design weakness that apparently made th,e circult susconfible to erratic operation if " contact resistance? between the 24 Vdc bus and the power supply monitor wire tb' develop, and the A developmentofahighresistanceconnec{ ion,(1.e.,abadcrimp I )
\' d connection) in the wiring betweeri ine4 24 Vdc bus and the power
, 3 xy supply monitor which exposed the distgr/ weakness and caused the module to trip. (SMUDhasagree[tofurtherexplorethecause g
of the failure of the power supply monitor by having an independent laboratory conduct ahditional analyses).
~
DISTRICT RESPONSE
~s s
Specific plant modifications were engineered and installed to correct w design weaknesses identified., Engineering Change Notices and subsequent field work acco gh ished the following:
? % ,w p
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~. - _
.-_...,. . .- . . _ . ..--.- -- - . - _ _ . . _ - . . _ _ = . - - . _ . - ._ _
i leads to the power supply monitor now the bus on the ICS and NNI. b g\@
Inspection and correction of terminations (i.e., lugs) was completed.
J I the power supply monitor has been sent to an independent !
b laboratory for analysis.
+
The District voluntarily undertook a program to test and inspect l j 1 terminations throughout the Bailey cabinets, regardless of systems.
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s 9-2
._~_._,___._,_.,,.-y,_,,,,, . . _ , . _ _ _ , , , ,,
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FINDING _9 J PBb O
$$@A#
f Upon loss of ICS dc 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 affected valves from the control room which necessitates t
manual actions locally at the valves.
DISTRICT RESPONSE i
Engineering Change Notices and subsequent field work have i
changed the configuration of valve positioning upon loss of ICS Power. Turbine Bypass and Atmospheric Oump valves now close automatically upon loss of ICS power.
l(
In addition, the AFW valves can now be manually controlled from the control room following loss of ICS power.
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! 9-3
__ . . . , _ . _ _ _ _ _ . - _ . . - , - - _ _ . , . , , , . , _ . . . - - , , _ , . . - -, ,..._._y _- .
l
- W 4, N !
, FINDING - 9.3
- g# s*# ' '
O 3.
+.
f#
An IFW manual isolation valve could not be shut by the 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.
It appears that the lack of a maintenance program could affect the operability of other manual valves at Rancho Seco.
DISTRICT RESPONSE O
Troubleshooting identified a lack of lubrication and rusted yoke nut bearings on Auxiliary Feedwater Isolation Valve FWS-063.
Reworking these components restored the valve to an operable condition. As an element of the troubleshooting effort, the
- similar valve cn the OTSG-B line (FWS-064) was inspected, as were all similar valves in service on the AFW system. All were found serviceable with only normal closing torque required to operate through their full travel, although evidence of recent lubrication was missing.
I In recognition of the desirability of having certain manual 4
valves readily operable, the Nuclear Operations Manager has identified a list of over 140 valves which will be verified
%/
9-4 i
b eV og a s&lfortoresumptionofpoweroperation. These manual 90 g lation valves will be characterized by their purpose and need
- to allow isolation of important active equipment such as pumps, valves, and heat exchangers. They will be selected to include both primary and secondary plant systems. Function, not class, will be the selection criteria. The program will involve actual stroking of the valve and, where necessary,. servicing with lubricants, packing, or adjustments. Statistics will be collected and evaluated to determine a summary status of valves in similar service.
l 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 and quality of the program. Specific procedures detailing the PMs are being expanded to provide confidence in the operability of the PM'd equipment. This expanded PM program will include the above
]
l identified valves in addition to those already receiving periodic maintenance, and any which meet the criteria being developed for this program.
i O
9-5 1
,m- - ,----.,, ,.- - - - , - - - - , - . - - - , - - - , , , _ - , , - , , - - - , , - < - . , , - , - - - -
,--.,,,,,-,--.,,-n,,,,,,.,,,,,_n,,c---+,
, , - , - - - - - - ,--- - , - , ne , ,_-
, FINDING - SECTIG% -
i g-b o Seco Emergency Operating Procedures (EOP) do not
! ~,5 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 I 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 i
not included in the Rancho Seco E0Ps.
OISTRICT RESPONSE The Emergency Operating Procedures have been revised and a Casualty Procedure unique to loss of ICS Power prepared.
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4 9-6 2
FINDING,c _
v f;i#
he E0Ps at Rancho Seco direct the operators to trip the e 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 stop the AFW pumps even when they had difficulty stopping flow to the once-through steam generators (OTSG) by valve 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 f to be the result of the substantial emphasis placed on the
\'
AFW system by NRC and others, and a lack of confidence in the reliability of the AFW pumps (i.e., fear that the pumps would not restart if stopped).
DISTRICT RESPONSE, The E0Ps did not contain specific parametric criteria for when to trip main and auxiliary feedwater pumps during an i overcooling, such as RCS Temperature or Steam Generator Level.
Lacking specific criteria, led the operators to be influenced by perceptions of NRC concerns and equipment reliability.
Procedures have been revised to specify when to trip main a auxiliary feedwater pumos. This eliminates an S$ o trip pumps when needed. j@
9-7
p FINDING %
V t,. Y' i t
$. $ T operators had considerable difficulty reconciling the dichotomy between avoiding the pressurized thermal shock (PTS) region (e.g, reducirg high pressure injection (HPI) flow] and regaining pressurizer level (e.g., increasing HPI flow in accordance with their ECPs). Their training and procedures were not adequate to resolve this conflict and to some extent tended to provided conflicting indications of the appropriate priorities.
DISTRICT RESPONSE
- 1. Emergency Operating Procedure (EOP) rules specifically state the D events or plant conditions which mandate throttling of HPI flows. Under Section 2.2 it is clear that HPI sho'ld u be throttled to prevent exceeding brittle fracture limitations.
- 2. The training programs for all License Training have been examined and conclude that, HPI throttling with no pressurizer level, is adequately addressed through the following!
- a. Both initial License Training and Senior License Training 4
m /
^ GY ProgramsaddressEOPrul n
J
- b. Many drill scenarios in the Simulator TraKdTfg gram contain the application and use of g s
9-8
_____. . _ . _ _ . - . _ . . _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ . . _ - . _ . . _ _ _ _ _ _ _ . . _ . _ . . _ . _ _ _ _ _ ~
1
- c. Agt es are required to be committed to memory by all 916b trol room operators and are frequently tested on both ip e NRC License and Rancho Seco Requalification Examinations ;
annually.
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$ b 9-9
O O
- [] FINDING 4 2 6 j b 'gp" h- The operators received neither classroom nor simulator training on the overall plant response to either the total loss of ICS dc power or the restoration of ICS dc power.
- DISTRICT RESPONSE Training programs have been revised. Operator simulator training has been increased by 607. for this year and will increase from one to two weeks for 1987.
l The simulator training included the following items:
t
)
- Emergency Operating Procedures (EOP) Training including all steps necessary to terminate overcooling or OTSG overfill from any cause, including the loss of ICS power.
l Changes to ADV, T8V, and AFW valve operation following loss of ICS power.
i Command and control training including Emergency Plan implementation, f t Recovery from SFAS, i .e. , restoring normal makeup and letd<gs flow. p f o -
9-10 i
t.
- cesbetweentheB&WSimul iC
( .
4
% g@ raps).
HPI and AFW throttling and trip criteria.
PTS recovery actions.
]
Cooldown rate interpretation and tracking.
i
, Conversion from AFW to MFW flow.
i The District is in the final stages of procuring its own plant
, specific simulator. Not only will the procurement of a simulator at Rancho Seco afford additional crew training, but the simulator will i also incorporate Control Room Design Review Human Factors l modifications. Thus, operator training will be current and will l reflect real-time plant conditions.
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9-11 <
f FINDING - 9. /
' \,
x b
- \ 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 1C bus and the ICS dc power supplies de-energized, the most credible cause for the loss of ICS dc power was the opening of switches Si and S2.
j However, the operators did not recognize this fact and, as a result, did not shut the switches untti 26 minutes into the transient. The fact that several operators did not recognize
- that switches Si and 52 were OFF suggests that their training on this crucial system was not adequate. In addition, although simpitfied drawings of the non-nuclear instrumentation (NNI) power supplies were posted on the NNI cabinets, comparable
'\
k drawings for the ICS power supply had not been provided.
DISTRICT RESPONSES Training has been conducted. In addition, Si and 52 labeling has g' been engineered and installed at the instrument location. A one b
ICS diagram has been posted in a conspicuous cabinet locatl . $4
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9-12
/ FINDING - 9.9 /
b 9 s not appear that nonlicensed operators properly operated
$t,, the AFW (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 determined the position of the valve while attempting to shut it completely. As a result of his actions, the valve was damaged, reopened, and the manual (local) capability to operate the valve was lost. These consequences suggest training weaknesses in the acceptable use of valve wrenches, the proper methods for manually operating and overriding air-operated valves, and the use of available and backup indications to determine valve positions. These weaknesses suggest areas where hands-on 6'
training ratner than walk-through or talk-through training may be necessary.
DISTRICT RESPONSE Hands on training is being given. Formal classroom and in-plant training on manual operation of these valves (and other power operated valves) has been underway. Specific training regarding -
/
operation of valves with valve wrenches has been given. Av wrenchguidanceStandingOrderhasbeenissuedandwill4e incorporated in procedures. s4'
/
v 9-13
FINDING - h4 N'
% \kilethedeficienciesinSMUD'sradiologicalcontroland
, t@f emergency preparedness programs during the December 26, 1985 incident did not jeopardize the public health and safety due to the relatively minor radiological consequences of this incident, they do indicate weaknesses in SMUD's program and the training of Rancho Seco personnel.
DISTRICT RESPONSES The District has reviewed its radiological control and emergency preparedness programs and made procedure and training improvements.
Recommendations for addltional changes are being evaluated as a part O of the PPIP.
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9-14 1
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,Q FIMDB 1 b %49
',N 11. The NRC staff was led to believe that the emergency feedwater initiation and control (EFIC) system would be installed in 1984 in response to a number of NRC requirements, including TMI Action Item II.E.1.2. Apparently SMUD decided to install an alternate system in response to II.E.1.2. SMUD's intent to satisfy II.E.1.2 with this alternate design was not made clear to the NRC staff, was not approved by the staff, and may not have complied with the requirements of II.E.1.2. As a result, the EFIC system, some features of which would have reduced the severity of the December 26, 1985 incident, has not yet been installed at Rancho Seco.
OV DISTRICT RESPONSE The AFW/EFIC (II.E.1.2) scope and schedule changes had been provided to the NRC staff.
The NRC issued Safety Evaluation Reports in January and September 1982, assuming EFIC installation. In October 1982, the District indicated that it would install interim safety grade AFW modifications and that EFIC was separate and beyond the AFW upgrade /
t requirements of NUREG-0737. The Olstrict also submitted a new schedule for EFIC implementation showing completion by Cycle 7.
schedule was confirmed by the Olstrict in December 1982. p i
v 9-15 u
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g 83, the District submitted a revised AFH system cription describing the interim AFH upgrades. NRR confirmed their
\@ understanding in an SER on the status of the AFH system dated t /
September 26, 1985.
I In November 1983, the District indicated that the interim system was l Installed thus completing Item II.E.1.2.
I Via a series of letters and living schedule submittals, the District J
informed the NRC that the EFIC installation was scheduled in two phases (Cycle 8 and Cycle 9).
i This understanding and approach was confirmed with NRR during a meeting in October 1985, at which time, the District committed to iI install the bulk of EFIC during The Cycle 8 outage.
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!U 9-16 k . , _ _ _ _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . . . _ _ _ _ _ _ _ _ . -
i FINDING - 9.12 %
g the RCS temperature dropped 180*F in 26 minutes, it i
' would have had to rapidly drop another 215'F (i.e., to an RCS temperature of about 170*F) while pressure was maintained at approximately 1400 psig, in order to seriously threaten reactor vessel integrity.
- DISTRICT RESPONSE The Olstrict agrees with this finding, based on calculations provided by B&W. In its letter (later) to the District dated (later), B&W also calculated the effect of the December 26, 1985, event on the reactor vessel as a function of the number of cooldowns consumed in 4
O the transient and the cumulative total of cooldowns versus the number I designed into the reactor vessel.
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9-17
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[ ,; FINDING-9.M
^ q $
p e 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 1
time this cccurs, 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.
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The Olstrict's PPIP specifically address six areas that are under full investigation to prevent trips to avoid challenging operators
, and/or safety systems, and to prevent a transient that causes the 4
post-trip response to leave acceptable pressure and temperature limits.
o
! 9-18
. _ _ _ _ _ . - - . ___ ._.___. ~ . _ _ _ _ . _ _ _ , . _ _ _ _ _ _ _ _ . . , _ _ . _ _ . . . _ . , , _ _ _
""I %
. s not clear that the overcooling transient was within the s Final Safety Analysis Report (FSAR) analysis of the Rancho Seco plant. Although PTS has been addressed generically, the FSAR accident analysts for Rancho Seco does not address this issue.
The most comparable analysis in the FSAR 1s for the cooldown due to a main stean line break. However, this analysis included only 100 seconds of the transient. In addition, the Rancho Seco FSAR analysis of main steam Itne 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.
O DISTRICT RESPONSE The Rancho Seco FSAR description of the main steam line break (MSLB) consists of several different analyses as shown below:
MSLB with ICS actions MSLB without ICS or operator actions MSLB with multiple stop valve failures The analysis without ICS or operator actions bounds the cooldown rate of the December 26, 1985, event. In this analysis, the cooldown rate is such that high pressure injection occurs in 23 seconds, followed by core flood tank injection 47 seconds after event initiation.
9-19
ption of the above analysts in the Updated Safety Analysis ort (USAR) is confusing and can lead to incorrect conclusions.
/ The analysis of the MSLB without ICS or operator actions was j
contained in the NRC question response section of the FSAR. The l analysis description was poorly incorporated into chapter 14 of the USAR. The District is currently revising the description of the MSLB analyses in the USAR for clarity. This clarification will be i
included in the annual USAR update in July 1986.
The subject of PTS was resolved by a rulemaking in December of 1985.
Prior to this time, interim submittals on the subject had been made.
l, PTS will be addressed in a future USAR update.
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! 9-20 i.
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i FINDING -
N
&#* ere were a number of precursors to the December 26, 1985, j , k incident at Rancho Seco...
I 4
I DISTRICT RESPONSE The District is committed to a permanent precursor review program.
J t
It is also expected that the Precursor Review, curreretly under way as
- part of the Plant Performance Improvement Program, will result in the identification of recommendations not only to address specific precursors, but also to determine whether the District's previous 1.
- analyses of precursors was too narrow in sccpe and worthy of i
additional action.
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--r.- , , - - -, y-+-,,. -,w#------ww-, - - - - - , -+---sv--
\ GN FINDINGS 48 s 4b[
4 M . It appears that the transient initiator (i.e., the loss of ICS
/
de pcwer) was not fully recognized by control room operators until two minutes after the power was lost. Although the "ICS and Fan Power Failure" alarm alerts operators about ICS power failures, it appears that its importance was somewhat obscured because it also acts as a trouble alarm for fan failure or for loss of one of the redundant ICS de power supplies, neither of which requires immediate operator actions or Initiates a transient.
DISTRICT RESPONSE
' The operators responded in accordance with the symptom-based E0Ps immediately, and had determined the loss of ICS power prior to the reactor trip, i.e., within the first 15 seconds. With regard to the annunciator alarms, Engineering Change Notices R-0517 and R-0580 were f issued to provide discrete window alarms for both ICS and NNI Power failures separate from any other alarms. This work is in progress i and will be completed prior to restart.
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\
l FINDING - 9.th b ,-
y f#
pt Annunciator Procedures Manual was not used by the operators i following the "ICS or Fan Power Failure" alarm. Even if the Annunciator Procedures Manual had been used, it contained very limited guidance concerning the Iniplications 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 others (EOPS, cps, etc.), has been revised. A long term procedure upgrade program includes all annunciator procedures.
i 9-23
1 I
FINOM eJh i b'~V G
44 /
[ 3. The ICS performance upon restoration of power is still not fully i understood, especially because performance may depend on the duration of the power interruption. However, when ICS de power i
1s 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.).
f 0! STRICT RESPONSE I
The B&W Owners Group is evaluating various aspects of the ICS l Including performance upon restoration of power. This evaluation
\ will consider the results of testing performed at Davis-Besse on the ICS performance upon power restoration.
The Olstrict has installed modifications which provide power and controls In the control rocm independent of ICS, which ensures that the demands from ICS during restoration of power will not cause a subsequent transient. Use of the controls has been incorporated into b ,,,
- the " Loss of ICS" procedure and training has been performed. og extent practicable, these procedures will be utilized 4(14 @
program. 4 D !
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4 9-24
, - - - - - - - - - , - . - . . - . . , . , - . ~ , . - - - ---,- _ ,-.,,., - _ _.- ..-- - , _ - . ,-, N
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FINDING e k
[\ q q, '@F t
$ Most of the indicators in the control room (both meters and 4
l recorders) are part of the NNI system; hence, they are generally l
- independent of the ICS. However, there are exceptions that had i
- not been recognized prior to the December 26, 1985 incident.
i j For example, the main feedwater (MFW) flow recorders are i
- affected by the ICS. During the December 26, 1985 incidsnt, the i
- recorder failed to a value near mid-scale when MFW flow was actually zero, i
)
l DISTRICT RESPONSE I
(later) 1 l
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FINDING t
$ B ause of a perceived sense of urgency, two nonlicensed 4
/ 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, procedures have been changed and training has been completed in response to the circumstances identified. Additional protective equipment, e.g., respirators, have been staged at
" locations more convenient to personnel requiring the use of this /
equipment. In addition, another health physics technician has bee added to each shift to provide dedicated H. P. support to Op Personnel. A s'
\
9-26
/ FINOINGcA.
m ,J 44 4 The operators did not remerrber a recent modtfication had been
' made to' permit the TBVs and ADVs to be closed from the remote i shutdown panel (outside the control room) Independent of the I
availability of ICS power. This change was made to acccmmodate I 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 j modification.
DISTRICT RESPONSE The Olstrict has installed modifications to allow closure of the TBVs U and ADVs independent of ICS from the control room. The use of the centrols has been incorporated into the operating procedures and training has been performed. In addition, modifications have been 9 5V installedsothattheTBVsandADVswillcloseuponlossofICSgc9 i
9-27
(
i
/
/ FINDINGofk
\ pv Additional staffing above that required by plant Technical 4 -
Specifications and other SMUD regulatory commitments allowed operators to perform certain tasks simultaneously. With 7 - staffing at the minimum required level, the actions performea I would have had to be performed sequentially, would have taken i longer, and could have exacerbated the overcooling transient.
/
) ; DISTRICT RESPONSE 4
,/
(
l [ The actions required to control the plant following the Icss of ICS j power event on December 26, 1985, could have been performed by the l D- minimum required staff. The District has implemented modifications
!\\-
and procedural changes which would decrease the demands upon the i
l operators should a similar event occur in the future.
! /
/
i i
4 l
O 9-28
~, _
4 N %
=,
FINDING - 9.23 $\ d'
? q@ - ,
p r heoperatorsnortheShittTshnicalAdvisor(STA)
, @ could identify an instance of when the STA provided engineering expertise during the incident. However, the operators found tiie STAvaluableasanextrapersondnshifttohelpoutduringthe incident. ,
DISTRICT RESPONSE Section 4.5 of NUREG-1195 indicates that the STA participated in the decision not to trip the Auxiliary Feedwater Pumps. This statement .
verifles that the STA did provide engineering input to the operators
! N decision-making process. However, the Olstrict is revie' wing the role
' ki of the STA in the decision-making process to ensure that the operators have engineering expertise available when needed. The i
specific actions to be taken are being dev>1oped and will be a l
l product of the District's Plant' Performance Improvement Program.
1 l
9-29
/'9 FINDING - t v ,y 4 appeared to the Team that SMUD personnel found the process of
/
troubleshooting in a highly controlled, systematic, and well-documented manner, as proposed by the Team, to be quite different from their usual maintenance practices. This difference. contributed to the difficulty that the Team experienced in reviewing the troubleshooting program.
DISTRICT RESPONSE Immediately following the 10/2/85 reactor trip, the District i
instituteo a systematic program for analyzing the event and resolving root causes. The program was based on NUREG-1154 Appendix B, which
- f\
described the Davis-Besse systematic troubleshooting program. The District's program was again implemented following a trip on 12/5/85.
't Following the 12/26/85 event, the District again implemented the i
systematic troubleshooting program. The project was in full effect when the IIT arrived on site. The IIT performed a line by line comparison of the District's program with NUREG-1154 Appendix B i
! without considering procedural and organizational differences between the two organizations. As a result, the District's plan was revised '
twice to incorporate IIT wording which in the District's opinto 4 not constitute substantive changes affecting the outcome o <$,h 4 l
troubleshooting. s (v .
9-30
g@ '
d T t had a highly controlled, systematic, and well documented
'(/ 1eshooting program in place prior to, and following the 12/26/85
' event.
i The root cause analysis for the 12/26/85 event was approved on March
{ 19, 1986, and issued on March 21, subsequent to the issue of i
' NUREG-1195.
In addition to establishment of root cause(s), the report identi specific areas for improvement. These improvements are ungh
/
\
/' '
1 l
)
2 I
9-31
- ~ . - _ - . -
O FINDI M s b ' qf ~
4 4, roughout 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 plans without including the actual data and analyses. As a result, the Team had to request the detailed underlying data and analyses, which subsequently were provided. This iterative process delayed the Team's on site investigation.
DISTRICT RESPONSE t' m The District's investigations following the December 26, 1985 event
.( )
A/ were broad in scope and exceeded that initially 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 withholding information.) As stated in Finding 10 when requested, the detailed Information was provided.
l 1
i The District did have difficulty in anticipating the areas where the IIT would desire detailed information. Significantly, the detailed f5 6 informationwasavailablewhenrequestedindicatingthattheDistri[ g
! had independently implemented an effective troubleshootin ,
o -
l 9-32 l
t'i FINDING - 9.2 $N V ,qf#
@ ne 1983, the B&W Owner's Group reported (BAW-1791) the
- results of an analysis which predicted an overcooling transient caused by a less 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 would occur approximately once every third year even if the EFIC system were installed at all B&W-designed plants. In addition, the report notes that one
.A 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, tne 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 District has implemented modifications and procedural changes s
which greatly decrease the likellhcod of future overcooling events.
In addition, further enhancements are being evaluated as part District's PPIP and the B&W Owner's Group STOP-TRIP c)b o
\ql 9-33 1
i i
f i !
$ . TENTATIVE & PRELIMINARY I _ F3 MSCUSSION PURPOSES ONLY -
I L
i APPENDIX A t
I l
j I
DISTRICT BOARD OF DIRECTORS' I
POLICY STATEMENT ON PERFORMANCE f.
6 '
IMPROVEMENT AT RANCHO SECO P
P f
i
{
f k
9 :
i
._,n. -___.,.,,.,,,nn_--,_ . , . _ , , , _ . , - -,__,.nc,, .,,.o._-.-,,
F i
i l
I l
, TENTATIVE & PRELIMINARY
_ pog gsCUSSION PURPOSES ON'Y -
1 I
I
- APPENDIX B l' ;
i COMPARISON TABLE ,,
DAVIS-BESSE, TMI, RANCHO SECO RESTART PROGRAMS (LATER) i l
l t
I i
i i
i I
h L
f
, , , . - - . , - _ . , _ _ . . - - _ _ _ . _ , - . . . - _ _ _ _ - . . . . . _ - . , , _ . . - , . . _ , , _ . - . . . _ _ _ _ _ _ . . _ _