ML20247N361
| ML20247N361 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 07/31/1989 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20247N334 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 PROC-890731, NUDOCS 8908020372 | |
| Download: ML20247N361 (106) | |
Text
p; a July 1989 '
Rev.2 Detailed Control Room Design Review I
{
Program Plan Pilgrim Station BOSTON '
EDISON RM"288?!888?86 F PNV
NUREG-0737, Item I.D.1 Generic Letter 32-33 BOSTCN EDISON Pdgnm Nuclear Pour staton Rocky Hdi Road P!jmouth, Massachusetts 02360 Ralph G. Bird BECo 89 112 senior vice Pres < dent - wclear July 24, 1989 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 License DPR-35 Docket 50-293 NUREG-0737, ITEM I.D.1: DETAILED CONTROL ROOM DE.SJEN_R_E_VJEW PROGRAM PLAN The attached report, " Detailed Control Room Design Review Program Plan (Revision 2)" addtesses the NUREG-0737, Supplement 1 (Generic Letter 82-33) requirement for the submittal of a program plan. Submittal of this revised Program Plan fulfills a commitment made in the "DCRDR Supplementary Summary Report" we submitted May 2, 1989, and describes our program for developing and implementing changes to the control room.
This Program Plan revision will be the basis for all Soston Edison DCRDR project activities from July 1, 1989 forward. Earlier revision of the Program Plan (Revision 0 and Revision 1) are no longer in effect. The earlier Program Plans are, however, valid for reference to the plans and procedures used in earlier phases of the project.
l The actions described in the enclosed report do not constitute changes to the Long Term Program. The report provides information on activities that are part l
of item No. 11, "CRDR-Phase III (Inventory, survey task analysis, selection of corrective actions)" in the Additional Items List, Attachment 1 to the Long Term Program Semi-Annual Report submitted by BECo letter 89-040, dated March 27, 1989.
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R. G. Bird PMK/jcp/3453 i
Attachment:
Detailed Control Room Design Review Program Plan cc: Oa next page l
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' July 1989 Rev.2 1
Detailed Control Room Design Review '
i Program Plan l
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i Pilgrim Station BOSTON EDISON 1
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5 TABLE OF CONT QIS I SECTION PAGE L Li s t o f . F i g u r e s . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -iv-
' List of Tables ..................................................... v ListLof Exhibits .................................................'. vi Preface ......:...................................... .............. vii SECTION I. : INTR 000CTION t1 A '. Purpose .................................................. I-1 B. Objectives ............................................... 1-1 C. Background ...............................................
I-1 D. DCRDR Eval uati on Cri teri a . . . . . . . . . . . . . . . . . . . . . . . . . . . . -. . . . I-2 E. : Acti vi ti es Si nce December 1984 . . . . . . . . . . . . . . . . . . . . . . . . . . . I F. Remaining Work ............... ........................... I-4
-G. Relationship to Earlier Program Plan ..................... I-5 H. Plant Description ........................................ I-6 I. Definition of' Control Ronm ............................... I-6 SECTION II. . PROJECT ORGANIZATION AND (QUALIFICATIONS A. Pil grim Station Organization . . . . . . . . . . . . . . . . . . . . . . . . . . . . II-1 B. Project Structure ........................................ II-1
'C. Key Individuals .......................................... II-6 SECTION III DATA COLLECTION AND ANALYSIS (HED IDENTIFICATION)
A. System Function and Task Analysis ........................ III-1 B. Comparison.of Display and Control Requirements to Control Room Inventory ................................ III-4 C. . Control Room Survey ...................................... III-8 D.- Operating Experi ence Revi ew . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . III-8 SECTION IV. ASSESSMENT OF HED'S A. Introduction ............................................. IV-1 B. Revised HED Assessment Process ........................... IV-1
.C. . Detailed Screening Process ............................... IV-7
'SECTION V. SELECTION AND IMPLEMENTATION OF CORRECTIVE ACTIONS A.- Introduction ............................................. V-1 B. Objectives................................................ V-1 C. Selection of Improvements ................................ V-2 D. Impl ementation of Design Improvements . . . . . . . . . . . . . . . . . . . . V-5 E. Veri fi cation and Clos e-Out of HED's . . . . . . . . . . . . . . . . . . . . . . V-6 11
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IAELE OF CONTENTS (Cont.) ;
1 SECTION PAGE SEC110N VI. COORDINATION HITH OTHER PROGRAMS A. General ..................... ............................ VI-1 8 Coordination Inherent in the Design Proces s . . . . . . . . . . . . . . VI-1 C. Coordination wi th Speci fic Other Projects . . . . . . -. . . . . . . . . . VI-1 D. Coordination Inherent in the DCRDR Project ............... VI-3 SECTION VII. RESPONSE TO NRC COMMENTS A. Introduction ............................................. VII-1 B. Safety Evaluation Comments ............................... VII-l C. In-Progress Audit (Inspection) Comments .................. VII-3 APPENDICES A. References ............................................... A-1 B. Resumes .................................................. B-1 lii
I LIST OF FIGURES ,
FIGURE NO. PAGE I-l Layout of Pilgrim Station Control Room . . . . . . . . . . . . . . . . . . . I-8 ]
II-l Boston Edison Nuclear Organization Chart ................. II-2 l
i l
II-2 Boston Edison Nuclear Engineering Organization Chart ..... II-3 V-1 Corrective Action Process ................................ V-4 iv L
LIST OF TABLES TABLE NO. PAGE I-l Compliance with DCRDR Evaluation Criteria ................ I-3 II-1 Technical Disciplines Assigned, by Task .................. II-4 II-2 Key Individuals Assigned to DCRDR or Providing Major Support .................................. II-7 III-1 Inventory Data ........................................... III-6 IV-1 Pre-Sc re eni ng Cat egori e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . IV-8 l
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LIST OF EXHIBITS
- l. EXHIBIT NO. PAGE IV-1 HE0 Documentation Form ................................... IV-3 IV-2 HE0 Assessment Questionnaire ............................ IV-5 IV-3 Classification of Plant Operations for l HED Screening ............. .............................. IV-10 IV-4 HED Importance Ratings ................................... IV-12 IV-5 Transient Event Categories for HED Screening . . . . . . . . . . . . . IV-13 V-1 Verification Data Sheet .................................. V-7 l
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I 1
l PREFACE Boston Edison submitted a Program Plan for the Pilgrim Station Detailed i Control Rocm Design Review Project (DCRDR) to NRC in October 1983, to comply j with the requirements of NUREG-0737, Supplement 1. After NRC review of the -
Program Plan, Revision 1 was prepared and issued in June 1984. In September l 1984, Boston Edison submitted its Summary Report to NRC.
After a pre-implementation audit in December 1984, NRC issued a Safety Evaluation in May 1985 in which NRC requested a number of improvements in the DCRDR processes, and requested a supplementary summary report.
Boston Edison submitted the Supplementary Summary Report to NRC on May 2, 1989. In that report, Boston Edison described progress since the previous report in 1984, identified additional work required to complete the DCRDR, and '
provided schedule commitments for the next phase of work. In that report, Boston Edison also committed to submit a revised Program Plan to update the detailed description of the scope and methods of the project.
This program Plan revision fulfills that commitment. It describes processes for the remainder of the Pilgrim Station DCRDR. This revision of the Program Plan replaces the earlier Program Plan for the purpose of work from July 1, 19P9 forward; the earlier revisions remain valid as the basis for past work done under those revisions.
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SECTION I INTRODUCTION A. PURPOSE The purpose of this updated program plan is to describe the methodology that Boston Edison will use to complete the Detailed Control Room Desiga Review (DCRDR) for its Pilgrim Nuclear Power Station.
In summary, this program plan will provide the following:
1
- Describe the work / tasks and the methodology that will be used to .
complete the DCRDR.
,
- Describe the DCRDR Assessment and Review Teams that will perform the DCRDR tasks.
B. OBJECTIVES The objectives of the DCRDR are as follows:
l
- Determine whether the control room provides the system status information, control capabilities, feedback, and analytical aids necessary for control room operators to accomplish their functions in an effective, safe and reliable manner.
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- Identify characteristics of the existing control room instrumentation, controls and other equipment, and physical arrangements that may affect optimum operator performance.
- Analyze and evaluate potential problems that could ari5e from this review.
- Define and put into effect a plan of action that applies human factors principles to enhance operator effectiveness. Particular emphasis will be placed on improvements affecting control room design and operator performance under abnormal or emergency conditions.
- Integrate the DCRDR review with other areas of human factors inquiries identified in the NRC Task Action Plan.
C. BACKGROUND Boston Edison originally prepared and submitted a Program Plan and a Summary Report (References 1* and 2) in 1984. The Program Plan described
- References are listed in Appendix A.
1-1
the project organization and methodology for performing a human factors )
review of the Pilgrim Station Control Room in accordance with the requirements of NUREG-0737 Supplement 1. The Summary Report described the work completed, listed the HED's identified during the review, and ;
outlined a series of correttive actions. '
NRC conducted a pre-implementation audit of the DCRDR program during the week of November 26-30, loN, and issued a Safety Evaluation Report (SER) I in May 1985 (Reference 3). The SER identified a number of deficiencies in l the DCRDR prograr and concieded that the corrective actions were not !
described in sufficient detail in Reference 1. NRC required that a supplemental report be prepared and submitted to resolve their concerns.
In August 1985, Boston Edison representatives met with NRC personnel at Bethesda, MD, to discuss the NRC's SER and Boston Edison's intended response. While the cccpe of the meeting included a number of topics, the central issue was NRC's concern with Boston Edison's approach to the System Function and Task Analysis (SFTA). The Technical Evaluation Report appended to the SER (Reference 3, pg. 11) concluded that substantial re-work would be required to produce an acceptable SFTA. In the August i
1985 meeting, NRC concluded that Boston Edison need only perform the upgraded SFTA process for the two new Emergency Operating Procedures (EOP's) that had not been issued or reviewed during the DCRDR process.
In April 1986, Boston Edison met with the NRC Project Manager at our offices and we committed to submit a Supplementary Summary report by a date four months after the end of the next refueling outage of Pilgrim Station (RF0 7). This commitment was documented in Reference 4; the Supplementary Summary P. sport was submitted on May 2, 1989 (Reference 5).
1 In March 1989, NRC conducted an in-process review of the Boston Edison DCRDR project (see Reference 6). Comments in that review are addressed herein.
D. DCRDR EVALUA_ TION CRITERIA Boston Edison recognizes and is responsive to each of the nine NUREG-0737 criteria by which the NRC evaluates DCRDP Program Plans and Final Summary Reports. Table I-1 identifies each of these evaluation criteria and the specific section(s) of this Program Plan revision or of the Supplementary Summary Report that describes compliance with each criterion for the Pilgrim Station DCRDR.
E. ACTIVITIES SINCE DECEMBER 1984 The recent Supplementary Summary Report (Reference 5) summarized the key activities on the Pilgrim DCRDR project since 1984. The report described physical improvements made as a result of DCRDR plus related activities that enhance the operators' ability to prevent and mitigate accidents.
l- The Report described additional corrective actions that have been selected to resolve some Human Engineering Discrepancies (HED's) as well as further engineering tasks to select corrective actions for additional HED's.
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TABLE 1-1 COMPLIANCE HITH DCRDR EVALUATION CRITERIA.
Cri teri a = Reference *
- 1. Establishment of a qualified- Section 11 mul.tidisciplinary review team.
- 2. Function'and task analyses section III A
-to identify control room operator. tasks and=information and controlt requirements during emergency operations.
- 3. Comparison of display and Section III B control requirements with 't
. control room inventory.
- 4. > Control room survey' to identify. Section III and deviations from accepted human Supplementary factors criteria Summary Report
'5. Assessment of HEDs to determine Section IV and which.HEDs are significant and Supplementary should be-corrected.- Summary Report
- 6. . Selection of design improvements. Section V and Supplementary Summary Report
- 7. Verification that selected Section V and design. improvements will Supplementary provide necessary correction. Summary Report
- 8. Verification that improvements Section V and will not introduce new HEDs. Supplementary Summary Report
- 9. Coordination of control room Section VI improvements with changes from other programs such as SPDS, operator training, l, Reg. Guide 1.97 instrumentation, and upgraded E0Ps.
- Section numbers refer to this P'rogram Plan revision; Supplementary Summary Report refers to Reference 5.
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j F. ,REMAININGJ_iQ8K .
This. Program Plan revision' describes'the methods for completing the DCRDR, The specif.c activities are summarized below for convenience, and
-described in more detail later in the Plan. The schedule fer these activities was provided in Reference 5 and will be updated as required via the Long Term Program process.
- 1. fvstem Function and Task Analysis l When the Pilgrim Station DCRDR hegan in early 1984, the symptombased Emergency Operating Procedures (E0P's) had been drafted but not yet 1ssued, and two E0P's had not been drafted. Boston' Edison committed to the NRC to perform additional System Function and Task Analysis (SFTA) on the two new E0P's when they were issued. The initial set of E0P's was based upon Revision 2 of the BWR Owners' Group Emergency Procedure Guidelines.
In 1987-88, the entire set of Pilgrim E0P's was rewritten and re-issued, including the t'<o E0P's not included in the 1984 SFTA.
The E0P's were upgraded te Revision 4 of the BWR Owners' Group Guidelines, which are substantially different from the Revision 2 guidelines. Changes in the E0P's are extensive. Boston Edison will perform a new SFTA on the entire set (rather than on only the additional two, as previously committed).
- 2. Vodate of the Control Room Inventory We will update the Control Room Inventory to reflect changes in the control panel configuration and panels added to the defined scope.
The inventory will be used in the Verification of Task Performance Capabilities to compare the information and control requirements determined in the SFTA to the actual control room configuration.
- 3. Verification of Task Performance Capabilities Following the completion of the SFTA, we will conduct a Verification of Task Performance Capabilities. The objective of this activity is to ensure the availability and suitability of the required control ,
room instrumentation and controls.
- 4. Validation of Control Room Functions He will conduct a Validation of Control Room Functions to determine if the functions allocated to the Control Room operating crew during I-mergencies can be accomplished effectively.
- 5. Control Room Survey He will conduct the control room surveys that were not previously completed. These surveys include:
- Noise Survey I-4
_ _ _ - _ _ - _ . _ - _ _ - - i
i.
Control Room Survey (cont)
- Heating / ventilation / air' cond,itioning (HVAC) survey Computer survey (Section 6.7 of.NUREG-0700)
- Sections of surveys 6.8 & 6.9 which were not done previously He.will also review all plant equipment that has been installed in
.the control room since the surveys were conducted in 1984. In
-addition,'we will survey the panels added to the scope of DCRDR since the previous survey. This survey will determine whether the new equipment conforms to_the NUREG-0700 guidelines.,
- 6. Assessment of HED's anti Selection of Corrective Actions The work previously performed plus the new work described above (items 1 through 5) will lead to preparation of a consolidated list of HED's. The HED's will be cssessed for significance as described in Section II of this Plan., He will then select design improvements or other corrective actions ta correct tLe HED's, as described in Section V._ (Many HED's wili atrehdy be addressed by the corrective actions described in Refer:tnu L)
- 7. Implementation of Corrective Att Qni 4 Selected design improvements and other corrective actions will next be planned and implemented. As part of the implementation Boston Edison will verify that the corrective actions will provide the necessary correction and will not introduce new human engineering discrepancies.
The results of the new data collection effort (items 1 through 5),
selection of additional corrective actions, and schedules for
. implementation will be reported to the NRC in a final DCRCR summary report, by November 30, 1990.
G. RELATIONSHIP TO EARLIER PROGRAM FLAN This Program Plan is intended to be the basis for all DCRDR project activities from July 1, 1989 forward. Earlier revisions of the Program Plan (Rev. O and Rev.1) are no longer in effect. . The earlier Program Plans are, however, valid for reference to the pl3ns and procedures used l in earlier phases of the project.
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H. PLANT DESCRIPTION The Pilgrim Nuclear Power Station is located on the western s'hore of Cape Cod Bay in'the Town of Plymouth, Plymouth County, Massachusetts. It is 38 miles southeast of Boston, Massachusetts. Bechtel Corporation was the architect / engineer and constructor of the station. The station consists of one 670 MW(e) (nominal) unit. It is powered by a single cycle, forced circulation General Electric Boiling Water Reactor producing steam for direct use in the General Electric 1,800 RPM tandem compound, four flow, non-reheat. turbine generator. Commercial operation of the unit began in December 1972.
I. DIDNITION OF CONTROL ROOM For.the purposes of the DCRDR project, the control room is defined as the following control panels, work stations and adjacent areas used by the operators for normal and emergency plant operations:
FRONT PANELS 903 Core Standby Cooling Systems 904 Reactor Auxiliary Systems 905 Reactor Control C-1 Feedwater & Condensate C-2 Turbine Control C-3 Electrical Systems C-76 Shift Supervisor's Console C-77 Reactor Operator's Console C-170/C-171 Post-Accident Monitoring CP-600 Augumented Off-Gas C-220, C-221, C-114, C-115 and Simplex CPU -- Fire alarm panels
- BACK PANELS 902 Area & Process Radiation Monitoring 910 Process Radiation Monitoring 911 Area Radiation Monitoring C-4 Feedwater Heaters Control i
- Panels to be addressed in second round of data collection, 1989-90.
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f l DEFfNITION OF CONTROL ROOM (Cont.)
C-7 Containment Ventilation & Gas Treatment C-174/175 Post-Accident Sampling
- The DCRDR maybe extended to other panels or areas in the control room if warranted by the analysis of selected events during the System Function and Task Analysis.
See Figure I-l for the idyout of the Pilgrim control room.
Panels to be addressed in second round of data collection, 1989-90.
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- 2. SOME DETAILS OMITTED FOR CLARITY.
- 3. NOT TO SCALE FIGURE I-1. LAYOUT OF PILGRIM STATICN CONTROL F<00M I-8
1 SECTION II PROJECT ORGANIZATION AND QUALIFICATIONS A. PILGRIM STATION ORGANIZATION The Boston Edison nuclear organization is illustrated by Figure II-1.
Responsibility for the Detailed Control Room Design Review Project is assigned.to the Nuclear Engineering Department for technical leadership and' project management. Considerable support to the project is provided by departments reporting .to the PNPS Station Director, including the k Nuclear Training Department and the Plant Department.
The ' engineering department organization is. shown in Figure II-2. A Project Manager, reporting to Department Management, has responsibility for coordinating.the DCRDR project.
~ Technical responsibility for-the processes, procedures and corrective actions rests with the line managers for the respective discipline groups in the Nuclear Engineering Department. Division managers assign personnel, either on a full-time basis or on a task by-task basis, to accomplish project tasks assigned by the project manager. In this fashion the normal Boston Edison engineering and construction processes for design control and coordination can.be used, thus minimizing the need for project-specific processes and controls.
B. PROJECT STRUCTURE Table II-l shows the key activities and the planned assignments of technical disciplines to each activity. As shown in the table, Boston Edison will-take the leadership roles in remaining tasks, but substantial support will be obtained from consultants, particularly for the inventory, survey, and SFTA activities. Resumes are in Appendix B.
Data collection will be conducted under the direction of personnel from the Control Systems Division (lead on inventory and survey updates) and the Systems and Safety Analysis Division (lead for Task Analysis and associated activities). Substantial consultant assistance will be
- required. The specific consultant is expected to be selected during the summer of 1989.
Design of physical corrective actions (i.e., plant design changes) will be under the direction of the Control Systems Division in most cases. The corrective action designs will be performed under the normal Boston Edison procedures for Pilgrim Station design cha ges.
Particular emphasis has been and will continue to be placed on the need for and value of substantial review and input from the Operations Section:
Operations has assigned a senior staff SR0 to act as liaison to DCRDR and help assure coordination.
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L NOTES TO TABLE 11-1
.l. All assignments for new data collection are tentative, pending selection of a consultant organization. Table excludes department management, division supervisors, the project manager, control room crews, and clerical personnel; and contractor management and clerical personnel.
- 2. BECo personnel may include a seconded contractor personnel working under BECo supervision and BECo QA program.
- 3. I/C means~ electrical, electronics, or controls engineers from'the BECo Nuclear Engineering Department Control Systems Division. Tbc irincipal Investigator (Harren Babcock, Jr.) is included in +M., category.
- 4. Systems engineer means a senior engineer from the NED Systems and Safety Antlysis Division. This discipline corresparids to the nuclear systems engineer in NUREG-0700.
- 5. Other design engineers means civil, mechanical, or electrical engineers assigne.1 to specific corrective action tasks (e.g., design of lighting improvements); or whose knowledge of existing design is needed for data collection.
- 6. Operations liaison is Mr. Ken Taylor, an SRO assigned to coordinate Operations Section involvement. Other control room operators are used as required in various tasks but are not assigned on a continuous basis.
- 7. Miscellaneous. support includes training department personnel for procedure revisions, training on design changes, and coordination with simulator for access and modification (personnel not specifically assigned to DCROR).
- 8. Other design specialists means contractor engineers delegated to design corrective actions (e.g., lighting improvements).
1 II-5 l
- J
m, t
-Corrective action designs have been and will continue to be assigned to contractors in most cases. Most panel-related improvements currently in design have been developed by Mr. Rett Considine of Management Analysis
! Co. He' has been involved with the Pilgrim DCRDR since 1984, and his L resume is enclosed (Appendix B).
Table II-2 lists the key individuals currently or recently assigned to the project; their resumes are in Appendix B. (Additional or different personnel .will be assigned if the project requires it.)
C. KEY INDIVIDUALS Key individuals on the project team and their responsibilities are as follows.
- 1. Proiect Manaaer David A. Bryant is the DCRDR Project Manager. He has served as the Project Manager of the DCRDR since 1985 and will continue to have overall management responsibility of the DCRDR. .He is responsible for the DCRDR budget, schedule, and inter-group coordination. He reports to the Manager of the Nuclear Engineering Department.
- 2. Princioal Investigator Warren Babcock Jr., is the Principal Investigator of the DCRDR. He has served as the DCRDR Principal Investigator since 1984 and will continue to serve as the technical lead for the DCRDR. He reports to
, the Control Systems Division Manager (in the Nuclear Engineering Dept.) and is responsible for understanding the applicable requirements (NUREG-0700) and applying them through the procedures described in this plan to produce appropriate corrective actions.
- Mr. Babcock is an experienced instrumentation engineer with substantial experience in panel design and additional training in human ~ factors engineering.
- 3. Lead Systems Enaineer The lead systems engineer currently assigned to the DCRDR is Mr. David Gerlits II. His principal responsibility is the new System Function and Task Analysis and associated tasks. Mr. Gerlits holds a Senior Reactor Operator's license, which will facilitate the integration of the operating aspects with the engineering aspects of the SFTA.
'4. Ooerations Liaison Mr. Ken Taylor is assigned as liaison between the engineering organization and the plant operations section for the DCRDR project.
Mr. Taylor holds a Senior Reactor Operator's license and has fifteen years of experience as an operator and watch engineer. He also has J been involved in the DCRDR project since 1984.
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l TABLE II-2 i
KEY INDIVIDUALS ASSIGNED TO DCRDR OR PROVIDING MAJOR SUPPORT l Position Individual Participation
)
Project Manager David A. Bryant
- Manages entire DCRDR project (full-time).
i Principal Harren Babcock, Jr.
- Lead technical person for DCRDR; Investigator member of Control Systems Division; ongoing department responsibility for human factors aspects of design process.
Sr. Systems David H. Gerlits, II
- Lead for systems engineering; Analysis Engineer responsible for planning and supervison of SFTA and related tasks.
Operations Kenneth N. Taylor
- Coordinates Operations Liaison participation in all phases; contact for information and i operator resources.
I/C Engineers Norman Eisenmann
- Responsible for various tasks, Anthony M. Horse, Jr.* including lead on survey and John V. Catalogna* inventory, support to SFTA; Robert M. Byrne lead for specific corrective action designs, support to other tasks.
Sr. Systems Kathleen D. Hard
- Lead for design of HED assess-Analysis Engineer ment process; support to various tasks.
Control Panel E. L. (Rett) Considine
- Integrated planning and design Designer (Management for corrective actions for all Analysis Co.) assigned control panel HED's.
Human Factors Danna M. Beith
- Review of proposed panel design Consultant (Management changes; planning of new data Analysis Co.) collection; verification of installed design changes.
Control Systems Siben Dasgupta e Supervises all Control Systems ,
Division Manager Division Engineers; responsible for technical quality of all I/C tasks; managt? ongoing HFE review of design changes initiated outside of DCRDR.
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TABLE Il-2 (cont'd)
Position Individual Participation Operations Section Leon Olivier
- Supervises all control room Manager operators and associated staff; provides guidance on DCRDR scope and direction, and resources for SFTA and related activities.
Simulator Thomas Beneduci e Coordination and support for all Division Manager activities in simulator.
Notes to Table
- 1. Only currently assigned key individuals are listed. Many other personnel have contributed or will contribute.
- 2. Table excludes contractor personnel (to be selected in Summer of 1989) for new data collection (SFTA, inventory, survey, etc.).
- 3. Any changes in assignments will be reported in the Final Summary Report.
- 4. Personnel designated with
- are seconded contractor personnel working I under BECo supervision and QA program.
- 5. See resumes in Appendix B.
II-8 1
SECT 10N III {
DATA COLLECTION AND ANALYSIS (HED IDENTIFICATION) l This'Section describes the new data collection and analysis effort to be {
conducted in 1989-1990, including updates-of the previous (1984) control room l inventory and survey, an entirely new System Function and Task Analysis, and comparison of ' display and control requirements to identify Human Engineering Di.screpancies (HED's).
Assessment of the resulting HED's is described in Section IV.
A. SYSTEM FUNCTION & TASK ANALYSIS
- 1. Introduction -
When the Boston Edison Detailed Control Room Design Review began in ,
early 1984,-the symptom-based Emergency Operating Procedures (EOP's) had been drafted but.not issued, and two additional E0Ps had not been ;
drafted. Edison committed to the NRC to perform additional System Function and Task Analysis (SFTA) on the two additional E0Ps when they were issued (see Reference 3). The initial set of E0Ps was based upon revision 2 of the BWR Owner's Group Emergency Procedures Guidelines (EPGs).
In 1987-1988, the entire set of Pilgrim E0Ps was rewritten and reissued, including the two E0Ps not included in the 1984 SFTA. The E0Ps were upgraded to revision 4 of the BWR Owners' Group EPGs, which are substantially different from the revision ? guidelines. The changes in the E0Ps are extensive. Boston Edison has decided to perform a new SFTA on the entire set of E0Ps.
The purpose of the SFTA is to determine the action and information requirements and performance criteria for the tasks that operators are required to accomplish under emergency conditions-as defined by the PNPS E0Ps and their associated satellite procedures. (Analysis of operator tasks associated with normal, abnormal, and alarm procedures will be excluded form this analysis, except where they
, lead the operator into emergency procedures). These requirements and the performance criteria will serve as the benchmarks for examination of the adequacy of control room equipment and instrumentation during the verification and validation activities.
- 2. Methodoloav Overview The SFTA methods and procedures will establish an objective, top-down approach to accomplish the following objectives:
a) Identification of the PNPS plant specific systems used for response to emergency conditions, j
1 III-l l
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= _ _ - - - _- )
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b) Identification of the operating events to be analyzed and their translation into functional requirements.
c) Detailed development of operator information and control requirements frcm the function descriptions.
Throughout the SFTA process, the emphasis will be on eventually identifying and analyzing operator action and information requirements for those tasks performed under emergency conditions.
The identification of event sequences and operator functions, the performance of function analysis, operator task identification, and task analysis will utilize expertise in systems engineering and analysis, human factors analysis, and plant operations. The process will be conducted independent of instrumentation and controls utilized in the control room. Human factors expertise will be heavily involved in all phases of the SFTA.
- 3. Detailed SFTA Methodology The following is a detailed description of each of the three objectives of the SFTA process:
a) Identification of the PNPS Plant-Soecific Systems and Subsystems -- The identification of PNPS systems and subsystems required for response to emergencies has essentially been performed as part of the Procedure Generation Package for the PNPS Plant-Specific E0Ps. i The PNPS E0Ps are based upon the Boiling Hater Reactor Owner's Group (BWROG) Emergency Procedure Guidelines (EPGs), which were developed in accordance with the requirements of NUREG-0737, item I.C.1. Revision 4 of the EPGs have been submitted to and approved by the NRC. These generic EPGs have been made plant-specific by the preparation of the PNPS Plant-Specific Technical Guidelines (PSTGs) and appendices of the E0Ps. These PSTGs served as the technical foundation for the development of the PNPS E0Ps. The development of the PSTGs included the substitution of plant-specific values for setpoints, design limits, etc., in the generic guidelines. The calculation of plant-specific limit curves and values was performed using the appropriate calculational methodology provided by Appendix C of the EPGs. Also where the EPGs detail the systems to be used, the appropriate or equivalent PNPS systems were specified.
b) Identification of the coeratina events to be analyzed and their translation into functional recuirementi -- BECo will use the PSTG as the function and system basis for conducting the SFTA.
The E0P Procedure Generation Package, particularly the PSTGs and E0Ps, define the PNPS systems and system functions. The allocation of functions to associated plant systems, and needed operator and plant equipment actions are all demonstrated in the I
III-2
7 structure of the E0Ps. Since the SFTA requires that all of the emergency actions performed by control room operators be analyzed, the E0Ps will be selected and used in determining the l operators' informational and control needs during the SFTA process.
c) Detailed development of ooerator information and control requirements from the function descriptions -- The initial step in the detailed development will be to separaie the functional requirements from the PSTGs, E0Ps, and satellite procedures into specific tasks. These will include all entry conditions l procedure steps, cautions and notes. Then, the analysts will determine the behavioral elements necessary to accomplish each task. This information will be compiled on " Element Tables".
The element tables summarize the basis for the step, and the requirements that the analysts will use in determining the action and information requirements for each task Included in these element tables will be Task initiation information requirements Task feedback information requirements Task 6acision requirements Task knowledge requirements Task action requirements Consequences of task error / omission These element tables will provide the baseline information for determining the action and information requirements for each step, The next step in the SFTA process will be to generate a list of action and information requirements for each task specified above. Confirmatory or alternate indications and controls that might be needed to confirm the performance of operator task steps will also be annotated on the data sheet.
Informational characteristics for the instrumentation will include parameter monitored, dynamic range, setpoints, precision (accuracy) at which the reading must be made (e.g., tS7.), speed of response to operation of associated controls, units, and the need for trending and alarming. Control characteristics will include the type (discrete or continuous). If the type of control needed is discrete, then the control characteristics such as detent versus spring-loaded, momentary contact positioning, and position (open-closed, stop-start, on, off, auto) will be specified, as well as feedback information associated with control.
This information will be tabulated and entered into a comprehensive data base. The data base will break down each task into behavioral elements. The behavioral elements will be defined by the action and informatic,nal requirements of the step, as well as any behavioral or physical properties required III-3 l
by the action. This will ensure that all of the properties of each behavioral element will be identified and captured.
At this point in the analysis, the information gathered will be in task sequence order, since the most effective method of analyzing operators' actions will be to place them in an accident scenario and allow them to respond. This will result in a series of action rad information requirements which are arranged in task sequence order. Numerous systems and their i associated information and control requirements would be '
interspersed through the sequence, making it very difficult to compare the information and control board inventory and to determine the presence and adequacy of controls and displays during the verification phase. In order to facilitate the verification phase, the information gathered in the database described above will be sorted from a task sequence o: der into a system-function-parameter order. This will be done for every system, subsystem, function, and parameter in the SFTA database, and a summary for each system will be generated. This summary will not add or delete any data, nor will it modify any of the information gatnered from the information and action requirements.
The results of the detailed analysis will be a database of operator action and information requirements. This database, along with the control room inventory data base, will be used as input into the verification of task performance capabilities.
This verification will assess the availability and suitability of instruments and equipment used by control room operators. In addition, the results of the SFTA will be used to assist in selecting event sequences for analysis during the validation of control room functions.
B. COMPARISON OF DISPLAY AND CONTROL REQUIREMENTS TO CONTROL ROOH INVENTORY
- 1. Uodate of tt1 Control Room Inventory The objective of the control room inventory is to develop a comprehensive listing of the instrumentation, controls, and equipment contained in the control room. This list is used in sutsequent tasks to determine the adequacy of control room components for supporting operator information and control requirements identified during the task analysis.
It should be noted that the control room inventory will be kept up to date, and will reflect componert changes made in the control room during the DCRDR.
Project personnel will conduct a systematic inspection and review of the control room and relevant control room documentation (e.g.,
instrument lists) to update the existing control room inventory.
III-4
The inventory records will contain the following information for each component (See also Table III-1):
- Instrument number / designation number t
- Component nomenclature or description
- System number
- Component characteristics (e.g., scale ranges)
- Panel No.
- Identification of Regulatory Guide 1.97 Instrumentation
- Other miscellaneous notes The product of the control room inventory is a comprehensive record of the instrumentation, controls, and equipment contained in the cintrol room. The control room inventory will be used in the verification of availability and suitability of the existing control room instrumentation.
- 2. Verification of Task Performance The objective of this activity will be to ensure the availability and suitability of required control room instrumentation and controls.
As recommended by NUREG-0700, this activity will be conducted in two parts: verification of availability, and verification of suitability. After completion of both the verification and validation activities, the problems identified will be documented as Human Engineering Observations (HEOs).
a) Verification of Availability -- The verification of availability will be done by comparing the operator information and control requirements identified during the task analysis with the control room inventory. The action and information requirements identified will be analyzed to determine the essential characteristics of control room instrumentation and controls that will be needed by the operator for acceptable task performance. This comparison will be performed on a component level, in order to verify the presence or absence of the required instruments and controls for each task analyzed during the SFTA. For any action or information requirement where an appropriate instrument, control, or other device cannot be found in the inventory, an HE0 will be generated.
b) Verification of Suitability -- Verification of suitability involves the examination of the human engineering characteristics of instrumentation and controls identified during the verification for availability. For this process, selected guidelines from NUREG-0700 and the criteria derived from the task analysis will be used. Such aspects of component design as the adequacy of the display range, usability of displayed values, adequacy of control type, adequacy and completeness of component labels, component location, and other characteristics which are unique to specific task sequences will be considered. Any deviations from the established criteria will be documented as HEOs.
III-5
i TABLE III-1 INVENTORY DATA'
- The .following data will' be- collected (as appropriate) for each component:
l Instrument Number The instrument number (e.g., PI3105) will be included to facilitate cross referencing between the inventory and drawings, etc.
Service Description
' A Functional description or the label will' be identified. The-information may be obtained from the pr.nels, P& ids, FSAR, Instrument Index, etc.
System NumSer The app:opilate system number will be listed for each item.
Manufacturer /Model The manufacturer and model will be listed when readily available. I Range / Units
" s range of each indicator and recorder will be recorded, switch positions will be recorded for each control switch, and lens colors will be recorded for each set of indicating lights.
The minimum scale increment / division will be recorded.
Panel Numbers The panel number where the instrument is located will be included.
Switch Tpe The switen type and handle type will be recorded.
Cross Reference Items that are part of a group will be cross-referenced to insure that all
. items within .a group.are reviewed together.
The' above-listed fields will be included in a computerized database as a m'nimum. Other fields may be included if they are determined to be beneficial.
III-6
- 3. Validation of Control Room Functions The objective of this activity will be to determine if the functions allocated to the control room operating crew during emergencies can be accomplished effectively within the structure of the E0Ps and their satellites, and the present design of the control room.
Emphasis will be placed on determining the adequacy of the control room design for supporting operator task sequences. In addition, the location of the required information and control characteristics will be considered, with particular attention paid to the unique characteristics of specific accident sequences.
The E0P validation sequences will be used, to ensure that all of the emergency tasks required by the E0Ps will be examined, to cover all of the systems in the E0Ps, and all the controls and displays used in the E0Ps.
The validation will be conducted by observing operators walking through_ the E0P validation sequences on the PNPS plant-specific simulator. The participants in the validation will be briefed concerning the objectives of the validation process, as well as initial plant conditions for the scenario being used.
The control room operators will be observed as they perform each sequence. As they perform the sequence, the operators will be prompted to describe their actions, including:
a) the cues by which they initiate a task b) the sources and adequacy of information (instruments, procedures, personal knowledge, etc.)
c) the application of information, including any mental conversions required, or uncertainties in the information provided d) controls selected and expected system response el methods for verifying system response and selection of alternative actions if response is not obtained f) indications that the sequence is proceeding hs expected g) indications that the sequence is complete h) other comments as appropriate The observers will record all relevant operator comments, as well as any observations that relate to the performance of the E0Ps or satellite procedures. They will also record:
- 1) any difficulties the operators have in responding to the scenario
- 2) the impact on operator performance of any previously identified HEOs or HEDs
- 3) any additional discrepancies identified during this task The results of the validation observations will be analyzed to identify any problems with the control room layout, obstructions to line of sight or operator movement, location of instrumentation and controls, operator workload, or other human engineering concerns.
These results will help assess the impact of previously identified i
III-7 l
I
___ J
9 HE0 and HEDs on actual operator performance. Any additional HE0s identified during the Validation process will be recorded and assessed in the same manner as the other HE0s.
C. CONTROL ROOM SURVEY A systematic survey of control room design features was started in 1984.
The portions of the survey that were completed in 1984 will be updated using the control room inventory as the base document. Changes to the inventory will trigger the update of the surveys. The surveys will be performed on checklists similar to those provided in NUREG-0700, Section 6.
In addition,.we will perform a survey update for each control panel as we complete installation of control panel enhancements. As part of the verification step after enhancements are installed, the completed panel will be reviewed.against NUREG-0700 criteria and checklists will be prepared. (See Reference 5 for description of the enhancements program.)
As was mentioned previo%1y in Reference 5, the computer survey (NUREG-0700, Section 6.7) and the noise survey (NUREG-0700, Section 6.1.5.5) will be performed after completion of the installation of the new plant computer and remon1 of the existing plant computer printers. Also, the HVAC Survey (NUREG-0700, Sections 6.1.5.1 and 6.1.5.2) can be completed only after certain maintenance and adjustments are made on the HVAC system. The surveys'from NUREG-0700. Sections 6.8 and 6.9, related to the SFTA will be completed after the SFTA is performed. (See Reference 5 for schedule information.)
Any discrepancies with human engineering principles will be documented via an HEO. Review and assessment of HE0's is discussed in Section IV of this Program Plan.
D. OPERATING EXPERIENCE REVIEH An Operating Experience Review (OER) will be conducted to insure that problems encountered in the operation of the PNPS main control room are addressed. The OER will be divided into two parts. One part will include review of available documentation of operating difficulties and incidents such as LERs that could be related to control room design issues.
Plant-specific data reviewed will be limited to the last five years (i.e.,
since the previous OER). NUREG-0700, Section 3.3.1 will be used as guidance.
The other part of the OER will include a Control Room Operating Personnel Survey. The survey will be designed to do:ument the knowledge that the control room personnel have of both the positive and negative aspects of the control room panels, environment, etc. The survey shall be made
-through the use of questionnaires and interviews. This is to identify factors that may help or hinder operator performance problems for examination later in the review. A range of operations personnel (Reactor Operators Sr. Reactor Operators, Shift Supervisors, etc.) will be interviewed to insure that as many different views as possible are captured.
III-8
ii.
' Interview sheets and questionnaires will be designed to gather information
, that can be used to determine appropriate changes to improve the.-
-operator's decision-making process. The following major topics will be
' addressed:
'. - Horkspace layout and environment.-
s -
Panel design
-Annunciator warning system Communications .
-- Computer system
- Corrective and preventive' maintenance Procedures Staffing and job design-
- - Training L
4The respondents to the survey and the personnel participating in the interviews will-be advised that they will remain anonymous.
The interviews will be structured to allow for additions of materials
, developed during the interview.
The data-developed during the interviews and surveys will be compiled to determine com:non areas of concern that need to be addressed as part of the DCRDR Project. Any items not meeting'the criteria of NUREG-0700 will be documented as Human Engineering Observations.
III-9
SECTION IV ASSESSMENT OF HED's f
A. INTRODUCTION In the Safety Evaluation Report (Reference 3) the NRC noted that implementation of the BECo assessment process, to determine which HED's are significant and should be corrected, had not yet been completed. This l section provides:an overview of the revised assessment process which has been applied to the original set of HED's', and will be used for future HED assessment.
After receipt of the NRC SER in 1985, Boston Edison started a process to improve the HED assessment process and used the improved process to re-screen the original HED's. The purpose of the improvements was to establish a more quantitative basis for setting priorities, to incorporate cost-effectiveness considerations into the process, and to provide better documentation of the HED screening criteria and results.
-The HED assessment procedure reflects the requirements of NUREG-0700, NUREG-0737 and NUREG-0800. In addition, it considers criteria for plant safety and availability.
B. REVISED HED ASSESSMENf PROCESS
- 1. Overview The revised assessment methodology supplements the HED assessment process described in the previously issued DCRDR Program Plan (Reference 1). The assessment process provides a method to categorize HED's for implementation, and it evaluates the potential consequences of unresolved HED's.
The overall process is as follows:
- Human Engineering Observations (HE0's) are identified and tabulated.
- The Human Engineering Observations (HE0's) are then assigned to Categories A, B, C, and D on the basis of their importance to plant safety, avt11 ability, and reliability.
- The appropriate DCRDR team members conduct the pre-screening process which separates all the HEDs into seven implementation groups for development of-the appropriate disposition.
s
- The team then conducts the detailed screening process which ,
provides a detailed analysis of specific HEDs to evaluate the i potential risk and the averted cost impact of the HEDs.
IV-1 i
Each of ~these major steps is described in the sections that follow.
The products generated by this assessment process are:
,. a. A list of HE0's grouped in four categories according to their safety significance.
- b. A list of HEDs grouped in seven categories, according to probable methods for implementation of corrective action,
- c. A list of HEDs ranked by relative importance to transient or accident risk, considering the risk of contributing to the initiation of an event as well as the potential to affect.
operator performance during response to an accident or transient.
- d. A list of HEDs ranked by relative importance to potential costs (costs th>t could result if the HED is not corrected, or costs that are avoided by correcting it). The cost impact considers four contributions: the potential for inadvertent plant scrams, equipment damage, extended outag: duration, and technical specification violations.
- e. Documentation that identifies the cost and risk impact from selected HEDs and the qualitative and quantitative bases for their relative importance rankings.
In summary, this procedure provides the estimate of the relative risk and the relative cost (lost productivity) but does not provide L
absolute measures for either quantity.
- 2. Identification of Human Enaineerina Observations Human Engineering Observations (HE0's) are identified in several ways. The principal source is in the DCRDR process during the control room survey and the SFTA. Additional HE0's have been identified during the course of the project by staff personnel and contractors while performing engineering tasks. Other HE0's are identified as a result of reviewing operating incidents (License Event Reports, e.g). Also, discrepancies identified by the NRC in the Pilgrim Control Room (see References 3 and 6) have been classified as HE0's.
For each HEO, a Form 1 is prepared (see Exhibit IV-1) that identifies the origin of the HEO, what guideline it does not conform to, the HE0's subject (panel, system), and other descriptive information.
HE0's are assigned a number and entered in a computer database.
. 3. Assessment of HE0's and Classjficejion as HED's by Safety Significance HE0's are then assessed for their safety significance and categorized by a specially-appointed Design Review Team (DRT) consisting of, at a I minimum, the following individuals: I IV-2 l
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- Systems and Safety Analysis engineer.
- Engineer with control room design experience.
- . Human factors engineer.
Additional team members may be assigned for all or only a portion of this procedure at the discretion of the project manager. A team leader is assigned by the project te administer the process.
The DRT reviews the entire HE0 as identified on Form 1, with the associated photo-documentation, if available. This,is followed by an open discussion to assure a. complete understanding of the discrepancy by all team members with the initiator present, as necessary. The human factors specialist participates and responds to questions from team members. During this process, the team may request s clarification of the HE0's under review. Any changes are indicated in the comment section of Form 1, with reference to'the reason for the changes.
The team then identifies the potential or actual operator error (s) and associated' safety consequences. Form 2 (see Exhibit IV-2) provides a format for the team for the systematic review of each HEO. The team leader completes a copy of Form 2 for each HEO. Any dissenting opinion is noted under the comment section of Form 2. A summary and reference to Form 2 is included on Form 1 under the section entitled, " Potential Operator Error (s)".
The team then determines which of the four afsessment categories (A through D) to assign to the HE0 under review and note the results on Form 1. The four categories consist of the following:
- a. Category A - HEOs Associated with Documented or High Potential Errors with Safety Consequence. This irrtudes HEOs which are known to have previously caused or contributed to an operating error as documented in a Licensee Event Report (LER) or other l
historical record, or as established by the interview responses of opt. rations personnel, cr which have the potential to cause an error of high safety consequences.
- b. Category B - HE0s Associated with Safety Considerations. This includes HEOs which have been determined by documentation or by potential to be of low safety consequence or to cause an unsafe condition.
- c. Category C - HEOs Associated with Availability or Reliability
,. considerations. This includes HEOs which have been determined i- to have potential for causing or contributing to human error that adversely affects the commercial aspect of electrical generating capabilities.
IV-4 e _ ._ __ _ _ _ . _ _ - - _ .____-____________m
+s NEDWI No. 392 Rev. No. 1 HE0 #.
FORM 2 - HE0 ASSESSMENT QUESTIONNAIRE COMMENT QUESTION YES NO (X)
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- 1. This discrepancy will cause undue operator faticue.
- 2. This discrepancy will cause operator confusion.
- 3. This discrepancy will cause operator discomfort.
- 4. This discrepancy presents a risk of injury to control room cersonnel.
- 5. This discrepancy will increase the operator's mental vorkload (for example, by requiring interpolation of values, remembering inconsis-tent or unconventional cont sl cositions. etc.).
- 6. This discrepancy will distract control room eersonnel from their duties. _
- 7. This discrepancy will affect the operator's ability to see or read accurately.
- 8. This discrepancy will affect the operator's ability to hear correctiv.
- 9. This discrepancy will degrade the operator's ability to communicate with others (either inside or outside the contr31 room).
- 10. This discrepancy will degrade the operator's ability to maniculate controls correctiv.
- 11. This discrepancy will cause a delay of necessary feedback to the coerator.
- 12. Because of this discrepancy the operator will not be provided with positive feedback about control tasks.
- 13. This discrepancy violates control room l conventions or oractices.
- 14. This discrepancy violates nuclear industry conventions.
- 15. This discrepancy violates population stereo-tvoes.
- 16. Operators have attempted to correct this discrepancy themselves (by self-training, temporary labels. " cheaters," " helper" controls, compensatory body movements. etc.).
- 17. Tasks in which this discrepancy is involved will be highly stressful (i.e.. highly time constrained. of serious consequence. etc.).
EXHIBIT IV-2. HE0 ASSESSMENT QUESTIONNAIRE (Pg. 1 of 2)
IV-5 f
l- NEDHI No. 392 Rev. No. 1 i-HE0 #
FORM 2 - HE0 ASSESSMENT QUESTIONNAIRE COMMENT OVESTION YES NO (X)
- 18. This discrepancy will lead to. inadvertent activation or deactivation of controls.
- 19. If this discrepancy caused a specific error, it is probable that another error of equal or more serious consequence will be committed.
- 20. This discrepancy is involved in a task which is usually performed concurrently with ancther task ( e.g., watching water level meter while manioulatina a throttle valve control).
- 21. This discrepancy involves controls or displays that are used by operator while executing emergency orocedures.
- 22. Assuming that this HED caused an operating crew error, it is likely that this error would result in:
- a. A violation of a technical specification, safety limit, or a limiting condition for operhtion.
- b. The unavailability of a safety-related system needed to mitigate transients or system needed to safelv shut down the olant.
- 23. This discrepancy involves controls or displays that are part of an engineered safety function or are associated with a reactor trio function.
Comments EXHIBIT IV-2 (Pg. 2 of 2) -
IV-6 j
[(
L d. Category D'- HEOs that are Minor or Non-Significant or that are l not to be acted upon (with justification).- This includes HEOs ll that have been evaluated and determined neither to increase the L potential for causing or. contributing to a human error nor to I-1.
have adverse safety consequences.
From this point on, items in Categories A, B, and C are considered to be Human Engineering' Discrepancies (HED's).
Note that individual HE0s may be reconsidered due to the cumulative or. interactive effects of multiple HEOs. Otherwise, HEOs (Category D only) could be discounted as non-significant and dropped out of the l assessment and improvement process. Effects of combined category-HEOs may be considered during selection of a disposition. Category D HE0s are not considered HEDs but may be corrected at Boston Edison's prerogative.
,4. Pre-screenina Process Performance of.the pre-screening process includes the following tasks:
e Review of each HED in the control room and verification of the
!iED against the configuration of the plant. Any changes to the as-built configuration and design documentation affecting the HED are noted as necessary.
- Each HED is assigned to one of the seven implementation categories ~in Table IV-l'. If applicable, any references or justifications are noted in defining the appropriate group chosen for a particular HED.
If during the review process or at any time after the completion of the review it was found necessary to transfer the HED to another category, the documentation is revised to properly track HED implementation.
C. DETAILED SCREENING PROCESS The detailed screening process provides a detailed analysis of HEDs in Category 1 (annunciator-related), Category 4 (hardware type), and Category 5 (hardware location).
These three categories were chosen for the detailed analysis because of their potential cost, complexity, and disruption to the Control Room in the implementation of corrective actions. For the other categories detailed analysis and prioritization is either unnecessary because corrective action for the items is justified without the detailed analysis, or the type of HED is not suited to this type of analysis.
The initial use of this detailed screening process was conducted at the Pilgrim control room mockup with the multi-disciplinary team present including operations and human factors representatives. (In future, the screening would be done at the simulator facility if feasible.)
IV-7
TABLE IV-1
, PRE-SCREENING CATEGORIES 4 ' l. Annunciator-related HEDs.
- 2. HED's for correction by surface enhancements (paint / label / tape / meter scale) or minor relocations of instruments, or switch handle change.
- 3. . Control room habitability and environment-related HEDs.
- 4. Hardware-rele.ted HEDs' associated with a less-than-desirable choice of equipment type or manufacturer based on human factors concerns.
(Example: improper shape or size of component or method of changing-componentistatus.)
5.. , Hardware-related HEDs associated with~a less-than-desirable location for the component relative to the operator's performance of normal or emergency procedure tasks using the component under review.
- 6. 'HED's that-are potentially resolved, pending verification and validation.
- 7. HED's-with administrative or operations disposition (Example:
operations procedure changes).
Source: Reference 7.
I IV-8 1_u_____________ ___ _ _-_ __
l.
l L, The detailed analysis is performed by a team using the approved ,
procedure (Reference 7). .To guide the team, the procedure includes forms and tables'of HED rating criteria guidelines..
The impact of- risk is determined through two types of contributions:
L (1) potential for-the HED to affect operator performance during their response to a plant ~ transient or accident, and (2) HEDs that can p contribute to the initiation of an event by affecting routine i operator performance during plant power operation, startup, shutdown,
! cold shutdown, or refueling. Qualitative evaluations of the relative significance of each HED are combined with quantitative information from representative probabilistic risk assessments (PRA) to evaluate the composite risk impact based on the frequency of possible operator errors and'the consequer.ces from those errors.
The impact of averted cost (i.e., potential cost if not corrected) is
' determined through four types of contributions: . potential for'the ,
HED to: (1) cause an inadvertent plant scram, (2) cause damage to l plant equipment, (3) cause unanticipated extensions to scheduled plant outages and (4) affect the operators' ability to maintain conditions within the limits set by the plant technical specifications. The averted cost impact evaluation follows the same format as the risk impact etaluation. Qualitative evaluations of the relative significance of each HED are combined with quantitative cost data to evaluate the composite cost impact based on the frequency of i possible operator errors and the consequences from those errors.
The detailed screening includes the following six tasks:
- 1. Identify the plant equipment included in the HED and determine all plant conditions during which this equipment is used from the following major categories:
- Equipment used during normal plant operation
- - Equipment used during plant startup/ shutdown
- Equipment used during plant cold shutdown / refueling p
- Equipment used during trasient/ accident response.
- 2. For each category selected from the above list, a separate fnrm will be completed for the equipment on each HED noting the following where applicable:
a) PNPS Operations procedure number applicable, b) System / function c) Type of Operation (see Exhibit IV-3) d) Frequency of use - Quantitative assignment from a low value of 1 for quarterly-annual usage to a high value of 5 for shiftly-daily use.
IV-9
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l NEDWI No. 344 l
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Revision 1 )
)
.1ABLE E. TYPES OF OPERATIONS DURING POWER OPERATION, STARTUP, SriUTDOW'i, COLD SHUTOOWN, AND REFUELING CONDITIONS Abbreviation Type of Operation 0 Observation, Monitoring, and Log Readings A Alarn Response and Problem Investigation C Manual Control, Alignment, and Adjustment T Testing M Maintenance x Otner EXHIBIT IV-3. CLASSIFICATION OF PLANT OPERATIONS FOR HED SCREENING IV-10
e )'- HED Potential- for' impact on Plant Operation for the.
1 following:
- Plant Scram o Equipment Damage m
h
- Technical specification violation e Loss of Shutdown Cooling
]
- Outage Extension-f) .Importance ranking is assigned to each HED based on the l factors listed in Exhibit IV-4. This importance factor is.
evaluated separately for each HED and each proced~re u step.
or occurence where the equipment is used.
? g) Time of equipment use during plant evolution
- Plant Startup/ Shutdown e Plant Outage
- Plant Transient
, h) Type of operator actions affected by.this HED from the following:
IDENTIFICATION - HED affects the ease with which the operators can use alarms, indications, displays, gauges, etc., to identify actual plant conditions and determine that response is required.
DIAGNOSIS & DECISION - HED affects the ease with which the operators can assimilate information from the available indications; use their experience, training and procedures to diagnose the causative factors; determine what response is required and decide to implement that response within the available time window.
RESPONSE - HED affects the ease with which the operators can physically implement the required response.
- 3. Develop quantitative risk impact indices for those HED's that affect plant transient response by performing the following:
a) Identify each scenario applicable to each HED from the listing i in Exhibit IV-5. This listing was developed from the plant-specific PRA performed on Pilgrim (Reference 10).
b) Determine the fractional risk contribution for each scenario from Reference 10 and similar units (sister plants), where applicable.
(Text continues on Pg. IV-18.)
IV-ll a_-______--____. . _ _ - _ _ _ _
NEDWI No. 344 Revision 1 TABLE 5. HED IMPORTANCE RAi!NGS Use the following numerical values to rate the importance of each unresolveo HED. If it 15 too difficult to Assign a single importance rating, indicate tne appropriate range of values on the HED evaluation form.
Not Very Low Low Mooerate ;,. Very High Inportant Ir:portance ;l Importance Importance Importance Importance 0 1 2 3 4 5
=
EXHIBIT IV-4. HED IMPORTANCE RATINGS l IV-12 l
l l
l L____-- _ _ _ _ _ _ - - _ _ _ _ - - - _ - _ - - - - - - - -_ --- -
i NEDW'l No. 344 Revision 'l
' TABLE 3. PILGRIM TRANSIENT EVENT CATEGORIES
. Abbreviation Transient c
GT' Broad Group: General Transient TTW8 Generator or Turbine Trip With Turbine Bypass Available TTOB Generator or Turbine Trip Without Turbine Bypass Available RXSC Reactor Scram FWPL' Partial Loss of Feedwater FWTL- Total Loss of Feedwater FWRU Feedwater Rampup MT0C Moderator temperature Decrease MSIl Closure of One Main Steam Isolation Valve MS!4 Closure of All Main Steam Isolation Valves PRFC . Pressure Regulator Failure Closed PriFO - Pressure Regulator Failure Open LOCV Loss of Condenser vacuum LODW Loss of Drywell Cooling LOHV Loss of Control Room HVAC
~LOIA Broad Group: Loss of Instrument Air LOSW Broad Group: Loss of Salt Service Water LOSP Broad Group: Loss os Offsite Power SL Broad Group: Small loss of Coolant Accident (LOCA)
SRV1 Spurious Opening of One or Two Safety Relief Valves RPSF Rer.irculation Pump Seal failure SLP8 Small Pipe Break LL Broad Group: Large LOCA SRV3 Spurious 0,ner.ing of Three.or More Safety Relief Valves LLPB Large Pipe Break ATWS Broad Group: Failure to. Scram O X Broad Group: Otner (specify)
EXHIBIT IV-5. TRANSIENT EVENT CATEGORIES FOR HED SCREENING IV-13
c). ' Calculate the unresolved HED difficulty index according to the following formula:
Difficulty Index - [3(I) + 5(D) + 2(R)1 (T) where:
. I - Ident5 fication Factor D - Dia;aosis Factor R'- Response Factor T = Tine Factor The numerical weights correspond to the relative importanc'e assigned for cognitive processes, information transfer and motor skills during operator response. The multiplicative time factor accounts for decreasing stress and improved error recovery as
-time increases after the. initiating event.
d) Calculate the Risk Impact Index according to the following-formula:
Risk Impact Index _(Fractional Risk Contribution)
X (Difficulty Index) e) Add the Risk impact Index values for all scenarios affected by this HED to calculute the total transient response Risk impact from the HED.
- 4. For HEDs which affect initiating events, the Risk Impact Index is calculated'as follows:
a) Identify the plant operating conditions during which this HED l
can cause an initiating event from the following:
- Normal Power Operation-
- Startup/ Shutdown
- Cold Shutdown / Refueling b) Identify the type of initiating events that can be caused by this HED from the listing provided in Exhibit IV-5.
L c) Determine the fractional risk contribution for each scenario L- from Reference 10 and similar units (sister plants) where applicable.
i d) Calculate the Risk Impact Index according to the following formula:
i Risk Impact Index - (Fractional Contributions) (Frequency l
Factor) X [10 (Importance Rating)).
-The multiplier of 10 for the importance rating corresponds to the sum of the weights for the identification factor, diagnosis factor and response factor in step 3 above. The multiplier is IV-14 m _ _ _ - - - - _ _- _ - -- - - - -
used because every initiating event challenges all three aspects of operator response.
e) Add the risk impact index values for all initiating event scenarios affected by this HED to calculate the total initiating event risk impact from the HED.
f) Add the transient response index and the initiating event index to calculate the total risk index for the NED.
S. Develop quantitative cost impact indices for those HEDs which affect any of the following cost areas:
- HEDs that can cause an inadvertent plant scram
- HEDs that can cause damage to plant equipment
- HEDs that can cause unanticipated extensions to plant outage schedules.
- HEDS that can contribute to the potential for technical specification violations The specific procedure steps applicable to each of the above are provided in the following sections.
- a. Cost Impact from HEDs affecting inadvertent plant scrams.
i) Identify the plant operating conditions during which the HED can cause a plant scram, from the following options:
- Normal Power Operation
- Startup/ Shutdown ii) Enter the appropriate condition frequency factor identified in Step 2d and the highest HED importance rating from Step 2f.
iii) Enter the approximate average cost per forced unit shutdown from the following cost data:
- Cost per forced shutdown from full power operations
- Cost per forced shutdown during plant startup iv) Calculate the cost impact index according to the following formula:
Cost Impact Index - (Frequency Factor) (Importance Rating)
X (Cost per scram)
- b. Cost Impact from HEDs affecting Equipment Damage.
Calculated using the same formula and factors as in 5a above, except with the cost per scram replaced by the equipment replacement cost for the specific equipment.
IV-15
~_ .
)
- c. . Cost Impact.from HEDs-affecting Outage Duration.
Calculated using the same formula and factors as.in 5a above,
.except with the cost per' scram replaced by the cost'per extended hou. at-cold shutdown multiplied by the number of hours of-
. extension anticipated.
-d.. Cost. Impact from HEDs affecting Technical Specification Violations.
4 -Calculated using the same formula and factors as in Sa above,
, except.with the cost per scram replaced with the estimated cost per technical specification violation.
- e. Add the plant scram, equipment damage, outage extension, and technical specification cost index values to calculate the total cost index for the HED.
.6.. . Rank the HEDs according to their risk and averted cost impact indices. These-indices provide the relative benefits resulting from the correction of the HEDs.. Thus, the risk indices and the averted
.. cost 1.ndices together provide a basis for ranking the HED's'according to the benefit of correcting them.
4 During the detailed screening process, HEDs (or portions of an HED) may be reassigned.to another implementation category for one of the following three reasons:
- Detailed investigation by the multi-disciplinary team identified that the corrective action has already been accomplished (verification may remain).
- The analysis.may identify the recommended reassignment. based on the present as-built configuration or the technical consensus of the multi-disciplinary team.
- Detailed analysis of the HED against the Pilgrim operating and l emergency operations procedures may identify no scenarios where
. utilization of the equipment identified in the HED would affect the risk'or averted cost in accordance with the criteria in the screening procedure.
The reassignment of an HED to another implementation category is documented with justification provided in accordance with the relevant procedure.
To complete the HED prioritization process the following activities are necessary:
i_
- Identify potential corrective actions and their potential l' implementation costs. Preliminary designs are first done for actions to correct the HED's subjected to detailed screening. The next step is to perform a cost estimate for implementation (engineering, construction, training).
IV-16
-.,a.-
rc i
- Compare benefits and correction costs to set priorities. Corrective action costs can be. compared to the HED risk and cost indices from Steps 4 and 5 to rank the HED's for cost-effectiveness. This ranking will be a major n wr in setting priorities for corrective actions.
Final selection of corrective actions will depend upon the HED rankings, the relative cost-effectiveness rankings, actual costs, ability to combine with other actions, and other engineering considerations (see Section V.C for further discussion).
q l
IV-17 l
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SECTION V. I SELECTION AND IMPLEMENTATION OF CORRECTIVE ACTIONS
.N.~ INTRODUCTION The' process by which each HED is assessed for resolution is defined in Section IV.- The results of the assessment are HEDs grouped in categories related to their most-likely resolution methods. At present, these.
categories are:
's Cat'egory 1:*' Annunciator-related HEDs~ l
- Category 2: Surface' enhancements HED's including:
Minor equipment relocations
' Minor equipment changes /replar:ements Improvements to meter and rectrder scales, mimics, demarcations, and labels
- Category 3: Habitability / environment HEDs
- Category 4: Hardware-related (significant equipment design) HEDs
- Category 5: Hardware-related (significant panel design) HEDs
- 2 Category 6: Potentially-resolved HEDs, including those justified for no action
- Category 7: Administratively-resolved HEDs The subject of this'section is groups of HEDs in the categories (1) through.(5) above. It will describe how improvements / resolutions are selected, tracked, reviewed, implemented, verified and closed out. HEDs in the categories'(6) and (7) are verified and closed-out only.
- The term " panel improvements" encompasses both improvements by " surface
. enhancement" (paint-label-tape), and improvements by other, more complex methods, called " design improvements."
- 8. OBJECTIVES
- The. objectives of this phase of the work are as follows:
- To insure that each HED selected for improvements (as discussed in Section IV), is resolved in the most cost-efficient manner consistent with the safety considerations of the HED.
' Category numbers are the same as defined in Table IV-1 but additional sub-categories are shown and category scopes are re-stated for this discussion.
V-1
p N 7 h
e That Human Factors Engineering (HFE) principles are used in the s' election of~ design improvements.
u e That;the proper. Human Factors guidelines are utilized in the detailed designs.for HED. improvements.
- That appropriate Boston Edison procedures are followed throughout the design and implementation process, e That the appropriate Operations personnel participate at all phases of.the design process.
- That a qualified Hvan Factors Engineer reviews the designs at selected points:
- That there is wrictu verification by the Human Factors Engineer that
-the design in fact resolves the HED and that no new HEDs are introduced into the control room.
J e That' cumulative and interactive effects of HED solutions are considered.
- That each HED is formally closed out in a uniform manner.
C. SELECTION OF IMPROVEMENTS Selection of improvements.can be divided into two categories:
- Analysis for correction by enhancement.
e' Analysis of alternative design improvements and selection of recommended solution.
L All HEDs selected for improvement will first be analyzed for resolution by enhancement. If' enhancement methods do not prove to offer a satisfactory resolution, more complex alternative design solutions will be considered.
It is intended that most proposed improvements receive a pre-installation review by control room operators. In some cases (e.g., control panel surface. enhancements and control room lighting) the proposed designs will first be in' stalled at the plant simulator to evaluate their effectiveness and to receive operator input. In other cases, review will be done via mockup.
- 1. Enhancement Methods Enhancements may be thought of as improvements which fall into the categories of labels / nameplates, mimics, demarcation, instrument scales, etc. These are often referred to as " paint-label-tape" improvements. The enhancements phase of the work is proceeding now at Pilgrim Station. Refer to the Supplementary Summary Report (Reference 5) for details of the HEDs currently assigned to enhancements and the corrective actions underway.
V-2
i 7 j As part of the enhancements work, four human. factors (HF) standards
~
have been developed to guide the development of improvements. .These j are:
- Nameplates & Labels
- Instrument Scales
- - Panel Demarcation
- Pan 01 Mimics Boston-Edison plans to consolidate.these four standards, plus others not-yet completed, into a "PNPS Control Room Design Manual," which will be used to guide future control room modifications.
-2. Selection of Desian Improvements f~
These corrections are those developed through planned design efforts. For- each HED, or group of HEDs, a preliminary conceptual design will be developed after consideration of various possible alternates. For some HEDs a formal design analysis report may be developed prior to selection of a conceptual design. Operations personnel will participate in the process of arriving at a conceptual design, as will the Human Factors Engineer and other applicable engineering disciplines.
.It is at this stage of the work that the possible effects of cumulative and interactive HED solutions will be considered. As is described in more detail below, we are developing design improvements on a planned basis, thus eliminating the deleterious effects of
" piecemeal" or isolated corrective methods.
Boston Edison has adopted the philosophy of grouping " design improvement" HEDs:(as opposed to " enhancement" HEDs) into two groups: " panel improvement" HEDs and_" global" HEDs. He believe this allows better planning, more efficiency, and a less-costly approach to' resolution of discrepancies than if each HED were treated individually. Records of'each HED will continue to be maintained on a separate basis, however. Refer to Figure V-1 for a flow chart of this process.
- a. Panel Improvement HEDI These HEDs lend themselves most efficiently to panel-by-panel resolution. Most design improvement HEDs will fall into this group.- '(It should be noted that many HEDs are applicable to more than one panel.) Conceptual designs for resolutions will be produced for each control panel within the defined control room. Various engineering disciplines, control room operators and the human factors engineer will participate in the design effort. Design documents will indicate which HEDs are addressed by that design, and their review will therefore be part of the V-3
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below.) Uniformity and consistency of design within one panel or across several panels will be assured by.use of a single.
design team, review of both conceptual and detailed designs by l the human facters engineer, and application of the PNPS human L factors. design standards.
The panel-by-panel concept has further advantages because it !
allows for most efficient allocation of manpower and capital, . >
'lexibility in implementation and minimization of disruption in tt.e control. room during construction.
b) Global Improvement HEDs Certain HEDs do not lend themselves to panel-by-panel
. improvements. These are related to discrepancies which must be' resolved on a control-room-wide basis, such as those involving l- annunciators, lighting, environment or workspace. For these
'HEDs it is planned that specific' studies be developed to censider the HED(.s), the effects of resolution, the types of resolution available and the costs for resolution, with recommendations as to the most prudent selection if more than one resolution is deemed acceptable. Engineering, operations and human factors personnel will participate in this effort.
Once a course of action is determined, conceptual designs will be prepared as discussed in the previous section. The first i stage of verification will also be initiated at this time.
D. IMPLEMENTATION OF DESIGN IMPROVEMENTS Only after conceptual designs
- have been approved by both engineering and operations can detailed design begin. . This is true for all types of improvements:- enhancements, panel improvements and global improvements.
It is at this point that the established Boston Edison procedures for plant design change (PDC) packages begin to take control of the concepts and they are treated more like any other plant modification and less like DCRDR work. The PDC packages for enhancements, panel-by-panel improvements and control-room-wide improvements will be assembled with all documents and instructions required to carry out the modifications. They :
will also carry all the specialized DCRDR documentation. The DCRDR '
Project will follow each package to keep track of its progress and to {
ensure continuity. Records of each HED and its resolution will continue to remain in the DCRDR Project, The DCRDR Project will require several l elements in each DCRDR PDC package:
- The term " conceptual design" is used here in a generic sense. In the Boston Edison PDC process, the approval of the conceptual design is via either the S.1A (Stope Justification and Approval form) or the conceptual PDC, depending upon the specific design change.
V-5 I
i j
l o; ' A design' report ' describing the various HF improvements contained in- I the PDC, how they were developed, and what they intend'to accomplish. j i
.* . Draft revisions to various DCRDR Project documents, such as HF ]
guidelines. checklists, task analyses, etc., when applicable' i e- Preliminary report of the project Human Factors Engineer stating that the design package fulfills the intent of resolving the HE3s contained therein.
- Initial verification report.
Once the PDC package ~is prepared, another detailed review by both the Engineering'and Operations departments is required before construction may begin.
E. VERIFICATION AND CLOSE-00T OF HEDS All HEDs selected for corrective action of whatever type must receive verification after completion of the corrective action. Verification is accomplished by meeting bath of the following criteria:
- The selected corrective action provides the necessary improv? ment.
The selected corrective action introduces no new HED into the control room.
Additionally, the following questions must be answered affirmatively:
- Has.the corrective action completed?
- Did'the completed. corrective action comply in all respects with the conceptual or preliminary corrective action contained in the HED records? (If not, was there a valid reason for the deviation?)
The final verification will be based upon as-built conditions (as verified Lby visual inspection of the finished item), to insure that the review includes effects of field changes or other variations from the design as originally reviewed.
The process of verification is controlled by NEDWI 392 (Reference 8),
Section 6.2.2. Refer to Figure V-2 for an example of the verification section. Each HED will be documented by this procedure. Initial documentation of verification will take place at the conceptual design stage for dose HEDs requiring enhancements or design improvements. This l ;is necessary to show that the conceptual design is of sufficient depth and l'
to show that the concept is satisfactory from a human factors point of view.
Close-out of each HED is documented via NEDHI 392, Section 6.2.3. Every HED, or HE0 with justification for no action, will be formally closed. By procedure, the verification must include approval by a human factors engineer.
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E SECTION VI COORDINATION HITH OTHER PROGRAMS A. GENERAL I Integration of Detailed Control Room Design Review (DCRDR) with other projects (and with other more routine activities) is important for at least two reasons. First, coordination helps to insure the ultimate success of DCRDR. Without it, projects can work at cross-purposes', with lower net benefits. -Second, coordination should reduce the risk that the control room, once improved by DCRDR, will be degraded (in a human factors sense) by'later activities.
Such coordination takes many forms, but generally the key is to utilize an engineering control process that operates across discipline boundaries and inherently exposes each group to the others' activities. This provides the mechanism for recognizing the need for and potential to achieve coordination.
There are three broad types of inter-project coordination activities:
those inherent in the design process, those inherent in the project itself, and explicit additional steps taken to help insure specific coordination.
B. COORDINATION INHERENT IN THE DESIGN PROCESS The Boston Edison design change process is multi-disciplinary, by procedure. Every Plant Design Change (PDC) involves review and concurrence by all engineering disciplines at the initiation of the design effort (via the review of the conceptual design and creation of a design input criteria sheet from each discipline) and when the detailed design change is issued. Each discipline reviews and approves the PDC and a multi-disciplined review is :anducted by a Design Review Board (DRB). The DRB has a formal charter and set of procedures. This multi-disciplinary review helps to identify and addre;s inter-project issues. In addition, all design changes require review and approval--again, at two separate steps--by'the Plant Department. Besides facilitating further inter-project coordination, this step helps to insure that operating concerns are incorporated in the design.
Another facet of inter-project coordination is that the Plant Design Change process itself utili7es checklists that help to insure that all relevent items are considered. In particular, the Control Systems Division checklist includes an item for NUREC-0700 review, when appropriate.
C. C00RDINA110N WITH SPECIFIC OTHER PROJECTS The DCRDR project explicitly coordinates with several projects.
VI-l
l.- Emeraency 091r3 tina Procedures-K The Emergency Operating Procedures (EOP's) are obviously central to the entire DCRDR. The focus of DCRDR is prevention and mitigation of accidents, as is the focus of the E0P's. The E0P's largely define the operating conditions and scope for the DCRDR affort.
Consequently, DCRDR is inherently connected with the E0P process.
The 1984 System Function and Task Analysis (SFTA) effort began with the original (draft) set of E0P's and the new SFTA to be performed
.(as. described in Section III)-will utilize the current E0P's, which are based upon the BWR Owners Group Revision 4 Emergency Procedure Guidelines.
Two other coordination steps occur. First, the Nuclear Engineering Department har ;he' technical cognizance for both the E0P's and'for the DCRDR. Throughout the recent (1987-88) effort to upgrade the E0P's, there was informal.coordilation betweEn the projects.
Potential Human Engineering Observations found through the-EOP development process were identified to the DCRDR team. In two specific cases the E0P project took action to add instruments to the control panels needed.for use of the Rev.4 E0P's.
Second, all Plant Design Changes are reviewed for their potential effect on E0P's. Thus, a design change for DCRDR would automatically be reviewed against the E0P's. (Similarly, all E0P changes must be reviewed for their impact on design.)
In-the new data collection effort described in Section III, the starting point for the new System Function and Task Analysis will be the latest- set of E0P's. As described there in more detail, the SFTA process is specifically designed to integrate E0P's with DCRDR.
- 2. Siffty Parameter Disolav System (SPDS)
As a major system designed specifically for use in the control room by operating personnel, the Safety Parameter Display System (SPDS) has been planned from its initiation as an integral part of the operating environment. The system design includes resolution of human engineering problems associated with the previous plant i computer system.
As one of the first corrective action projects initiated by DCRDR, new operator's and supervisor's consoles were designed and installed in the control room. The new plant computer was an integral part of the design, with provisions for EPIC /SPDS terminals for the operator and supervisor. !
Once the computer installation is completed, a DCRDR survey will be performed as described in Section III, which will determine if there are any unresolved human factors problems.
VI-2 i
l 1
_--_ ._ - -- - _ - a
l Future _' activities- that will further tie the two projects include'the l Task Analysis, which will inherently include the SPDS (to the extent i: that the SPDS is used in the E0P's). Also, an annunciator study is-to be done (see Reference.5 Section IV) that will consider the EPIC /SPDS computer as part of the integrated' alarm information i- sources, i
- 3. Reculatory Guide 1.97 As specifically required by Regulatory Guide 1.97, all appropriate control room devices will be marked to identify that they are qualified for the post-accident envi'onment.
The devices added to the panel by the RG 1.97 review will receive the
.same human engineering review as any other panel design change. In addition, the control room survey and inventory updates will automatically review the added devices for their conformance with Human Factors Engineering (HFE) guidance.
The DCRDR project also will coordinate with the RG 1.97 effort in several other ways:
a) DCRDR will check instruments for conformance to prescribed ranges from RG 1.97 as part of the SFTA verification step (comparison of information and control net Js to actual conditions).
b) The DCRDR control room inventory will designate RG 1.97 devices.
l 1: c) Panel changes (e.g., device replacements) identified by DCRDR will include consideration of RG 1.97 requirements.
D. COORDINATION INHERENT IN THE DCRDR PROJECT Because the DCRDR project will include a survey of the entire control room and will use operating procedures explicitly, the project will automatically come into contact with other control room-related personnel, design changes in progress, and hardware. Although this is a passive type of coordination, it acts as a check on the more explicit coordination I efforts just described. I
!- One coordination advantage in the project's organization is that there is no separate project workforce; the technical contributors are part of the same engineering groups performing all other projects and design )
modifications. Consequently, the project team members are continually and routinely exposed, through their work group, to informal information of other activities with a potential connection to DCRDR. The Division Managers responsible for those groups and the Section Managers to whom they report are thus common points of contact for all projects, helping to ensure coordination.
I l
VI-3
SECT 10N VII RESPONSE TO NRC COMMENTS A. 1RTR000CTION This.Section responds to NRC's comments on the original Program Plan (Reference 1) as contained in the Safety Evaluation (Reference' 3). In addition, it responds to NRC comments made during the March 1989 inspection and' reported in Reference 6.
Please note that many of the SER comments were discussed in the Supplementary Summary Report, Reference 5.
B. SAFETY EVALUATION COMMENTS 1.. Establishment of a Oualified Multidisciolinary Review Team The= Safety Evaluation concluded that Boston Edison had met the requirement for a qualified _ team. This Program Plan. updates the relevant information. (See Section II.)
- 2. System Function and Task Analysis The Safety Evaluation discussed.at length the previous SFTA conducted by Boston Edison. Boston Edison is preparing to perform a second SFTA, as described in detail in Section III. He consider that the new SFTA will be fully in compliance with NRC requirements.
The key items for. improving.the SFTA were cited on page 11 of Reference 3. Those-four items and references to the current description in Section III of this Program Plan are as follows:
a) " Include two additional guidelines from Revision 3 of the EPG's." -- This is resolved by using E0P's that were developed from Revision 4 of the EPG's which does include the two additional guidelines (secondary containment control and radioactivity release control). Please see Section III.A.3.
b) " Include a ' front-end' task analysis for all the EGP's including those~in Revision'3 ..." -- The front-end task analysis will be performed, as described in Section III.A.3.
c) " Include operator interview techniques and questions designed to identify SFTA information and control needs during the
' front-end' phase, and task analysis HED's during the walk-through phase." -- Control room operators will be used in the detailed development of operator information and control requirements, as described in Section III.A.3.c. The analysis will complete the " element tables" based on the information they receive during table-top scenario discussions with the control room operators. In addition, operator feedback will be used in VII-l
3 the Validation of Control Room functions (see Section III.B.3) and a-separate operating experience review'(see Section III.D)
.will be conducted.
d) " Expand the scope of the task analysis to include all operator
- tasks during all emergency operations, including going to hot
' shutdown." -- As stated in Section III.B.3, we will include all the emergency tr ss required by the E0P's and all satellite procedures referenced in the E0Ps. Appendix B of the Emergency Procedure Guidelines clearly states that one of the purposes of the E0Ps is to shut down the reactor and bring the reactor to cold shutdown conditions.. Therefore, the SFTA will include all emergency operations and will include going to hot shutdown.
- 3. Comoarison of Disolav and Control Requirements with Control Room Inventorv
' The NRC found the previous inventory process to be sat' factory; the previous inventory will be built upon and updated, as described in Section III.B.l. The comparison with inventory and display needs will be entirely re-done, as described in Section III.B.2.
- 4. Control Room Survey The.recent Supplementary Summary Report (Reference 5 Section V) included a detailed discussion of each NRC comment on the survey.
Please.also refer to Section III.C of this document for a description of the survey updating effort.
- 5. Operatina Exoerience Review Reference 5 includes a detailed response to NRC's comments.Section III.D describes the new Operating Experience Review to be performed.
- 6. Assessment of HED's to Determine Which' are Significant and Should be Corrected The Safety Evaluation found the original assessment process to be acceptable. Revision to the assessment process were described in Reference 5, together with results, and are described in more detail in.Section IV of this Program Plan. Application of the assessment process to HED's found during the next phase of data collection will be described in the Final Summary Report due to be submitted to NRC in Noveuber 1990.
- 7. Selection of Desian Improvements The methodology described earlier was satisfactory, according to the Safety Evaluation. NRC asked for additional information on how cumulative and interactive effects of HED solutions are to be considered, and for the results.
VII-2
Please refer to Section V'of this report for a more complete description of the design improvements selection process. Please refer to Reference 5 for the design improvements selected to date.
Remaining design improvements will be described in the Final Summary Report,
- h. VgIlfication that Improvements will Provide the NecessArv Correction without Introducing New HED's The Safety Evaluation requested a process description. Please refer to Section V.C.
- 9. Coordination of Control Room Improvements with Chanaes Resultina from
'Other NUREG-0737 Imorovemer,t Proarams The Safety Evaluation conclu' fed that coordination with the SPDS and Reg. Guide 1.97 efforts appeared to be adequate, but that coordination with the E0P development effort needed improvement.
This issue was discussed at length in the inspection reported in Reference 6 and in Reference 5. For further description of our coordination efforts, please refer to Section VI of this Program Plan.
- 10. Proposed Schedules Project schedules are presented in Reference 5 and will be maintained via the Long Term Program process.
C. In-Prooress Audit (Insoection) Comments These sub-sections discusses NRC's comments in Reference /,, reporting on the in-progress audit (inspection) of March 20-22, 1989 In general, Reference 6 summarizes NRC guidance for DCRDR's with many specific points mentioned. Most of NRC's concerns are addressed elsewhere in this Program Plan (or in Reference 5). Boston Edison does wish to respond here, however, to selected individual NRC comments which might otherwise not be readily resolved by the text in other Sections.
- 1. Prioritization of HED's according to Safety Significance (Page 9 of Reference 6)
NRC states, "It was not clear whether or how these two indexes [of the risk and consequences of operator errors associated with HED's]
will be combined to achieve a single prioritization of HEDs according to safety significance.... The licensee should ... provide assurance that safety considerations, not the cost of corrective actions, will drive decisions about which HED's should be corrected."
NRC is correct that the Nuclear Engineering Department procedure (Reference 7) for prioritizing HED's does not clearly establish a
" cut-off" for corrective actions, based upon their index scores. The indices provide two separate indications of the importance of the VII-3
HED, and thus of the benefit from its correction. Since the scales are dimensionless, the two indices will not be summed or otherwise combined.
Cost is another factor. Given two items of equal safety significance. BECo would give higher priority to the item that produces more benefit for a unit of resource invested -- that is, the one with the higher benefit-to-cost ratio would be done sooner.
Ultimately the selection of projects for resource allocation is a management activity, not a DCRDR project activity. Using the NEDWI 344 techniques, the DCRDR project will identify the benefit indices for the HED's evaluated, and the relative benefit vs cost for the groups of HED's, but Boston Edison management will ultimately make a judgment as to what corrective action to implement, and when, using several other considerations.
One consideration will be how best to coordinate related corrective actions, or HED corrections with other control room activities. As an example, two or more corrective actions might be combined to produce a logical construction package (in terms of cost, schedules, disruption, training, etc.) even though one or more HED's of low or intermediate priority were resolved "early" in that way. Also, the consideration of cumulative and interactive effects may result in additional corrective actions for a given design package.
As another example, an HED with very high safety benefit might be partially resolved (e.g., by enhancements) even though another aspect of it is deferred for schedule reasons.
In summary, the NEDHI 344 process is an aid to management decision-making; it provides additional guidance on the logical priorities for resolving HED's; it does not remove the need for balanced management judgment on the scheduling of work. In particular, it does not inherently resolve priority conflicts between the complex HED's analyzed per the NEDWI and simpler HED's nat assessed by that method; nor does it inherently indicate the relative priority of DCRDR work versus other projects.
Safety considerations will thus be explicitly considered in making corrective action decisions, but will be weighed along with other issues in the management decision process.
- 2. "In the upcoming supplemental summary report, the licensee should identify the correction status of each HED documented in the previously submitted summary report (Reference [2])."
The supplementary summary report (Reference 5) was prepared before we received Reference 6. Our report included information on the status of each HED, although the specific HED categorization scheme requested by NRC was not included. BECo will provide a final tabulation of HED's in the final report due in NRC November 1990.
- 3. "
...[T]he licensee does not have a formal process for verifying that ,
the proposed improvements will not introduce new HED's when implemented." (pg. 12)
VII-4
Boston Edison's formal process for verification that design changes do not intro /uce new HED's is summarized in Section V. The two primary elements are:
a) The design process includes a step in which a NUREG.0700 checklist can be specified, so that creation of new HED's can be identified before detailed design proceeds on any panel modification.
b) The HED cicsecut precess (Reference 8) includes explicit reviews of HED corrective actions so that new HED's, if any, can be identified and corrected.
I VII- 5
i
)
APPENDIX A F_Ef.EBf'1ES A-1
APPENDIX A REFERENCES
~
(1) _ . Detailed Control Room Design Review; Program Plan; June 1984, Rev. 1; Boston' Edison Co.
(2) Detailed Control Room Design Review; Executive Summary Report; Doc. No.
BEC0/ESR-1, September 1984, Rev.1; Boston Edison Co.
(3): Safety Evaluation by the Office of Nuclear Reactor. Regulation of the
- - ' Detailed Control Room Design Review for Pilgrim Nuclear Power Station, Docket No. 50-293; forwarded by NRC letter dated May 16, 1985 (D. B.
Vassallo to W. D. Harrington)
(4)- . Boston Edison letter to NRC BECo 87-008 dated January 20, 1987 (5) Detailed Control Room Design Review; Supplementary Summary Report; April
-1989, Boston Edison Co.; forwarded to NRC by letter BECo 89-064 dated May 2, 1989 (6) In-Pro'gress Audit Report of the Detailed Control Room Design Review at Boston Edison Company's Pilgrim Nuclear Power Station, dated April 12, 1989; forwarded by NRC letter to Boston Edison dated April 26, 1989.
(7) Boston Edison Company Nuclear Engineering Department Work Instru'tica No. 344, Revision 1, dated May 18, 1989. " Assessment of Human Engineering Discrepancies."
(8) Boston Edison Company Nuclear Engineering Department Work Instruction No. 392,-Revision 1, dated December 12, 1988. " Process for Documentation of New Human Engineering Discrepancies (HED) and Verification of Design and Implementation Completion for Correction of HED's" l (9) U. S. Nuclear Regulatory Commission, Standard Review Plan, NUREG 0800, Revision 1, Section 18.0 and 18.1 (with Appendix A), September 1984.
(10) BECo Probabilistic Safety Assessment, dated March,1988 (PLG-0616, Volume 1-4) s I
A-2 l
_ - _ _ _ _ _ _ _ _ . A
g 6! -
r APPENDIX B RESUMES
)
I I,'
B-1
n L.' ~' i NARREN BAB00CK, JR.
SR. ELECTRONICS ENGINEER BOSTON LDISON COMPANY PRINCIPAL INVESTIGATOR
' EDUCATION-Bachelor of Science, Electrical Engineering, Brown University,1968 Graduate Study, Industrial Engineering, Ohio University PROFESSIONAL REGISTRATION Control Systems Engineer, State of California PROFESSIONAL AFFILIATIONS Institute of Electrical and Electronics Engineers Human Factors Society PROFESSIONAL TRAINING Training in ' Human Factors Engineering:
Massachusetts Institute of Technology - 1980
" Man-Machine Interfacing" General Electric Nuclear Training Center - 1980 "BWR Owners' Group Human Factors Engineering Workshop" University of Wisconsin - 1981
" Human Performance and Nuclear Safety" PROFESSIONAL EXPERIENCE j!oiton Edison Company (1979 - Present)
Sr. Electronics Engineer, Control Systems Olvision, Nuclear Engineering Departmeat
}e Currently working as cognizant engineer for Oontrol Eoom Design Review 7; Project. . Acted at team leader of a BWR Owners' Group control room scrvey l-:. team, hember. '8VROG Cortrol Room Improvements Sub-committee. Also re.rpi.isible for design of new control systems and modifications to existing contr01. systems at Pilgrim Nuclear Power Station, including prtparatior, of-instructions for' installation of new equipment and procedures for check-out and testing of this equipment. Have served as instructor for operetor training ir, elettritallelectronic syst2m3
. Operation.
B-2
Harren Babcock, Jr.
- Page 2 r
EXPERIENCE (Continued)
Burns and Roe. Inc. (1977 - 1979)
'Sr. Engineer / Group Supervisor, Instrument and Control Department, Breeder Reactor Division Supervision of I&C engineering group with responsibility for design of balance-of-plant I&C systems for a breeder reactor project. Lead engineer, solid-state logic systems' design. Lead engineer, electronic security systems.
Ebasco Services. Inc.
Sr. Instrument & Control Engineer (1974 - 1977)
Designed'I&C systems-for application to nuclear and fossil power plants.
Reviewed vendor system design documents for compatibility with clients' specifications. Member of engineering team charged with design and layout.
responsibilities for control rooms at various power plants, both fossil.
and nuclear.
Crvoaenic Technoloov. Inc.
Electrical Engineer (1974)
Designed control panels and control systems for nuclear power applications.. Prepared field test procedures for documentation of installed system performance. Field engineer for checkout and testing or radioactive waste process systems.
Stone & Webster Enaineerina Corporation Control Systems Engineer Designed control panels and control systems for nuclear power applications. Prepared field test procedures for documentation of installed system performance. Field engineer for checkout and testing of radioactive waste process systems.
Begock & Wilcov Capjtn.y j
Electrical Eng'ineer, Nuclear Power Generation Dgartront f Designed and/or specified electronic control systens for nuclear steam supply systems when built in D&W plants. 11eviewed vendor specifications q and documentation for systems built outside B&H. Iratructed customers' q engineering personnel on operation and maintenance of B&H's systems. i i
B-3
0 DANNA M. BEITH
-PROJECT HUMAN FACTORS SPECIALIST F' HUMAN FACTORS INTERFACES
+
'(SUBCONTRACTED TO MANAGEMENT ANALYSIS CO.)
EDUCATION B.A., Psychology, University of California, 1976
. PROFESSIONAL AFFILIATIONS Human Factors Society-Associate Editor. Human Factors Society Bulletin PROFESSIONAL EXPERIENCE Human Factors Interfaces (1986 - Present)
President, Responsible for the management and direction of a consulting firm specializing'in human factors engineering and research, and nuclear support services.
Brunswick Nuclear Power Pl'nt a - Served as the human factors specialist for the verification and validation of the Rev. 4 Emergency Operating Procedures (EOPs). Duties included a detailed human factors review of the procedures and flow charts for logical flow, wording and consistency.
Also updated the E0P Writers Guide and Users Guide.
Conducted the human factors validation of operator performance on updated E0Ps and ERFIS displays. Duties included the development and implementation of the Team Operations Performance and Procedure Evaluation (TOPPE), which was developed to assess operator performance and acceptance of procedures / operator aids; the observation of operator actions on the plant simulator; and operator interviews at the completion of simulator scenarios and data analysis.
Developed a-Verification and Validation process for the evaluation of the Alternative. Safe. Shutdown Procedures. Duties included the training of operators on the process for the walkdown of the procedures, incorporating operator coments, ud the human factors review of the prccadcres.
Wrote the human factors sections of the Verification and Validation
. procedure for E0P changes /modificati,ons. The procedure ensures that human factors principles are considered witn each procedure modification.
l Participated at the Frar.n factors spec 1111tts in the SPDS display l -
development process. Duties included the development of displays, the evaluation of the'huma.n-interface requirem2nts and ensured the J compatibility of displayed information with the E0Ps. Alsr, conducted the human factors review of the SPD5.
1 B-4 l.
1.
l
_ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ _ . - - - I
F Danna M. Beith
- Page 2 l EXPERIENCE (Continued)
I H.B. Robinson Nuclear Power Plant - Currently developing a symptom-based flow chart for the Emergency Action Levels for the classification of emergencies. Duties include updating of Plant Emergency Procedures, developing.an operator training package, and EAL User's Guide, and an EAL Hrlter's. Guide.
Planned and conducted a human factors review of operator performance on updated E0Ps._ Duties included observations of operator actions on the plant simulator, operator interviews, data analysis, preparation of the-
- final report, and assistance in the resolution of problems identified with procedures and flow paths.
Conducted a human factors review of the Dedicated Shutdown Procedures.
Duties included detailed review of the procedure format, wording, and consistency between procedures; the rewriting / reformatting of the procedures; incorporating operator comments; and the preparation of the final report.
Conducted a human factors review of ERFIS/SPDS. Duties included the evaluation of the human-intv face requirements, system usability and compatibility with the E0Ps.
Conducted a human factors review of Maintenance Test Procedures. Duties included an analysis of trips related to these procedures, review of procedure usability / format and suggestions for improvements in the implementation of the procedure, and enhancements to the control panels involved.
Pilarim Nuclear Power Station - Conducting a human factors review of the control room design review control panel modifications for BECo's Pilgrim Nuclear Power Plant. Assist in the planning for the completion of the !
DCRDR and update the Pilgrim DCRDR Program Plan and Final Summary Report for resubmittal to the NRC.
b Lirl.Xt.in_P_ojAubit One - Conducting the human factors validation of operator performance and acceptance of the ECPs. Duties include the l implementation of the TOPPE, which was developed to assess operator J J performance and acceptance of proc (dures/ operator aids; the observation of ]
operator actions on the plant simulator; merator intervf ews at the ]
completion of simulator scenarios; and data analysis and preparation of 1 interim and final reports. !
l
$tteron_lfg.tli,lyclear Powar Plet - Conducted a human factors review of l the Emergency Action Level procedures and flow paths for the Shearon Harris Nuclear Power Plant. Duties included a detailed review of format, wording and consistency with E0Ps.
B-5
Danna M.'Beith Page 3
. EXPERIENCE (Continued)
'RMS Associates. Inc. (1984 - 1986)
Director. Human Factors Services - Managed the (NUREG-0700)' Control Room Design Review for Carolina Power and Light Company at the H.B. Robinson, Brunswick, and Shearon Harris nuclear power plants. Duties included task analysis, verification and validation, SPDS reviews, control room surveys, Human Engineering Discrepancies (HED) evaluation, preparation of final report, and assistance in implementation of control room modifications.
Wrote the program plan for the operating plant and the final summary report for all three plants. Developed a data base for system function task analysis which incorporated owners' group guidelines.
Essex Corporation (1980 - 1984)
Staff Scientist - Participated in the Control Room Design Review for Virginia Electric Power Company at North Anna and Surry Units 1 and 2 nuclear power plants. Conducted an operating experience review which consisted of writing operator questionnaires, interviewing operators, data reduction, and a document review of plant documentation, such as License Event Reports. Assisted in writing the VEPC0 program plan and photographing the control panel photo mosaics.
Research Scientist - Directed the on-site data collection for Toledo Edison's Control Room Design Review for the Davis-Besse Nuclear Power Station. Duties included review of operating experience, conduct of control room surveys, SPDS review, and a Human factors review of upgraded E0Ps. Assisted the photographing and construction of a control panel photo mosaic, data reduction, and preparation of final report.
Derformed the Human Factors evaluation of the South Texas Project Main Control Panel and Control Room for Bechtel/ Houston Lighting and Power (subcontract through Torrey Pines Technology). Activities included an evaluation of a full-scale, three dimensional mock-up prior to fabrication of the operational system and the set-up of a computer program for sorting and reporting data.
Project Manager for the development and production of approximately 300 nuclear power plant surveillance / test procedures for South Carolina Electric and Gas (SCE&G). work involved technical review and editing of procedures, technical direction for all project staff, and coordination of procedures production from initial writing through word procesr.ing.
fasponsible for technical staff of six to eight tcchnical writers, two editors, two nuclear power plant operations specialists, and eight word L prcces' sors. l i
B-6
-__________-_-_______-__-_D
l U
Danna M. Beith Page'4.
EXPERIENCE (Continued)
- On-site supervisor. for the rewriting / reformatting of nuclear power plant emergency, normal and standard operating procedures at SCE&G's Virgil Summer Nuclear Station.
,c Directed Human Factors evaluation of the on-site data collection for Comanche Peak Nuclear Power Plant control room. Evaluation included-criteria specified in NUREG/CR-1580 and NUREG-0700. Duties also included documenting and identifying human engineering discrepancies and backfits.
Research Associate - Participated in the (NUREG/CR-1580) Human Factors evaluation of three nuclear power plants for Carolina Power and Light.
One plant evaluation included a control board assessment of engineering drawings for a plant under construction. . Evaluated procedures developed for control room review; identified, reported, and suggested suitable backfits for human engineering discrepancies found-in the control room.
XEROX CORPORATION (1978 - 1980)
Associate Human Factors Desioner - Supported Human Factors Department in the Business Machine and Copier / Duplication Divisions. Duties included control system design, behavioral. testing, and new product assessments.
Also. wrote machine operating procedures and developed dialogues used for operator assistance.
CANYON RESEARCH GROUP. INC. (1978)
Assistant. Researcher - Contract research assistant to Xerox corporation, Industrial Design / Human Factors Department, Business Machines Division.
Duties consisted of control system design and behavioral testing.
B-7
t
! THOMAS BENEDUCI l.
SIMULATOR DIVISION MANAGER BOSTON EDISON COMPANY EDUCATION Associate Degree, Electrical Engineering, Franklin Institute of Boston, 1975 Bachelor of Science, Industrial Technology, Northeastern University, 1986 PROFESSION TRAINING Nuclear. systems training course designed specifically for Pilgrim Station, including specific studies on RHR, Core Spray, HPCI, RCIC, TIP, Neutron Monitoring SBLC, Turbine Generator and Reactor Vessel intervals.
School (five weeks) on BWR 4 Nuclear' Instrumentation including studies on the APRM, IRM, SRM, TIP, Area Rad Monitor, Log Rad Monitcr and Process Rad Monitor systems.
PROFESSIONAL EXPERIENCE Bostonldjson Comoany (1980 - Present)
Simulator Division Manager (April 1989 - Present)
Responsibilities encompass overall operation, maintenance and modification of the Simulator Complex. This includes management of the Simulator capital and expense budgets and varying number of management, union and contractor personnel in the planning and scheduling of Simulator modifications, discrepancy corrections, and enhancements. Manages or participates on special teams in analysis of plant transients tasked with root cause corrective actions being identified, initiated and completed.
Special projects include installation of a redundant Simulator computer system, installation of the Simulator EPIC computer system, installation of Simulator emergency preparedness phone systems and Simulator NRC Certification. Active member of the Utility Simulator Users Group Executive Committee, Secretary of the NETA Simulators Advisory Committee and a member of the Employees Speakers Bureau.
Sr. Simulator Hardware Engineer (October 1987 - March 1989)
Responsibilities included managemcnt of the following: the Simulator capital and expense budgets, hardware modifications for the SEP, emergency lighting HPCI Vacuum Breaker, other minor modifications, the computer room humidifier upgrade, and completing the Simulator spare parts inventory. Asr.isted in and conducted Simulator tours for media sno special interest groups.
Special projects were the research and approval of a tAckup Simulatial i Computer System, managing the instr:llation of the EPIC computer system at the Simulator, and writing purchase specifications for the Simulator EPIC System and Toshiba Intelligent Display Terminal relocation.
Industry-related activities included serving as secretary to the NETA Simulators Advisory Committee and being an active member of the Employee Speakers Bureau.
B-8 i
-Thomas Beneduci Page 2
. EXPERIENCE (Continued)
Sr. Modifications Engineer, Pilgrim Nuclear Power Station (November 1985 -
October 1987)
Responsibilities included administering post-construction acceptance testing of all Plant Design Changes (PDC), hiring and directing contract personnel for the administration / coordination of post-construction testing, and providing interface with-the plant maintenance, engineering, operations, and all other departments involved in the PDC process.
Special activities included providing reports to Senior BECo Management and team leadership on CAL 86-10 resolution, presenting positions to NRC on questions relating to CAL 86-10 and other testing / PDC issues, and acting as Modifications Management Group Leader during the absence of the group leader.
Instrumentation and Controls Supervisor, Pilgrim Nuclear Power Station (November 1983 - 1985)
Responsibilities included direct supervision of I&C Technicians and Contractors, scheduling personnel, writing and reviewing plant procedures, heting as project mananer for Plant Design Change packages, and directing installation of new plant equipment. This position required interfacing with other station groups including Nuclear Engineering for planning and implementing Plant Design Changes.
Instrumentation and Controls Technician, Pilgrim Nuclear Power Station, (August 1980 - November 1983)
Responsibilities included performance of maintenance activities on all categories of nuclear power plant equipment, writing and performing surveillance tests and postwork tests required to prove equipment performance meets technical specification criteria and operability requirements. Special projects included installation of the Seawater Differential Temperature modification, the new CRD Temperature Recorder, the Drywell Hi-Rad Monitors and various other plant design changes. Also responsible for writing and performing system logic tests required to satisfy NRC Bulletin 80-06 concerns.
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{
d DAVID A. BRYANT-
, PROJECT MANAGER b BOSTON EDISON COMPANY L DCRDR PROJECT MANAGER EDUCATION Bachelor of Science, Electrical Engineering. Tufts University,1966 MSE, Catholic University of America, 1971 PROFESSIONAL AFFILIATION L
Institute of Electrical and Electronics Engineers PROFESSIONAL TRAINING Certificate in nuclear power plant engineering program, Bettis Reactor Engineering School, 1967 PROFESSIONAL EXPERIENCE Boston Edison Company (1976 - Present)
Project Manager, Control Room Design Review Project, Nuclear Engineering Department _(1985 - Present)
Responsible for overall management of project, including assignment of tasks to project personnel (in-house and contractors); coordint. tion of efforts by all involved departments; administration of purchase orders; monitoring of progress and developing corrective action as needed; budgeting, scheduling and planning; and review and approval of licensing submittals and other correspondence.
Pilgrim 2 Project Manager (1981-1984)
Responsible to manage efforts to'close out cancelled nuclear power plant project, including negotiation of settlements for about 100 cancelled contracts; maintenance, marketing, and eventual disposal of $100 million of hardware; and liaison with regulators, twelve joint owner utilities, and various company organizations.
Project Engineer, Pilgrim 2 Project (1976-1981)
Project engineer and contract administrator for Nuclear Steam Supply
.Systent contract for 1100' MH PWR. Coordinated the review and approval of contractor design documents. Administered NSSS contract (approximately
$100 n.lllion). Managed interface among architect-engineer, NSSS supplier, and utility stafL on technical issnes, scheduling, and contractual taatters. i 1
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,- , Dave Bryant Page'2
- EXPERIENCE (Continued)
- l. i GCA/Technoloav Division (1974 - 1976)
Senior Engineer and then Group Leader Transportation and Land Use
. Planning Department Performed and managed environmental studies, transportation planning, and .i land use planning projects for government clients.
Fav. Soofford ik Thorndike. Inc. (1971 - 1974)
Transportation Engineer / Planner (
Performed traffic engineering, transportation planning, land use studies, l~ and environmental studies for public agencies. Performed field surveys, data collection, analysis, and report preparation.
U.S. Navy (1966 - 1971) i L
Nuclear Power Engineer, Naval Reactors Division, U.S. Atomic Energy Commission Naval officer serving in navy headquarters engineering organization as cognizant engineer supervising contractor efforts for design, procurement, modification and repair of mechanical equipment in nuclear ships.
I j
B-11
L ll ROBERT M. BYRNE INSTRUMENTATION & CONTROL ENGINEER
, BOSTON EDISON CO.
EDUCATION B.S., Marine Engineering, Massachusetts Maritime Academy,1977 LICENSES Chief. Engineer of Motor Vessels of no more than 7000 hp, Mineral and Oil Industry Third Assistant Engineer--Hotor Vessels--any horsepower.
Third Assistant Engineer--Steam Vessels--any horsepower.
PROFESSIONAL TRAINING
" Applied Human Factors in Power Plant Design and Operation," General .
Physics Corp., 1987 EXPERIENCE Boston Edison Company (Aug.1987 - Present)
Instrument & Control Enginet.- Nuclear Engineering Department, Control Systems Division.
Responsible for providing engir.'ering support to Pigrim Nuclear Power Station through designing, analyO ng and modifying I&C systems and components. Other duties include ps acaring Safety Evaluations and procurement documents, drawing reviews and providing engineering support to other disciplines within the nuclear organization. Familiar with NRC Regulatory Guides and IEEE Standards. Assigned to the DCRDR Project as cognizant engineer for Annunciator Conceptual Design study.
Stone & Webster Enaineerina Corporation (1980 - 1987)
Instrument & Control Systems / Turnover Coordinator Control Systems Division, Beaver Valley II Project.
Responsible for ensuring completion of engineering and design of control systems prior to transfer to Dusquesne Light Company. Acted as liaison among engineering (SWEC, DLC, and Site), construction, and operations to en6ure testing and start up of plant. Developed engineering analysis reports to management utilizing the Lotus 123 Software.
. Control Systems Engineer / Change Management Coordinator Responsible for reviewing potential changes of control systems for f'easibili ty. Evaluated construction impact of change to system.
Instrument & Control Engineer Responsible for developing, revising and reviewing logic diagrams and system descriptions. Prepared and reviewed controls systems section of the Final Safety Analysis Report for BVII.
B-12
Robert M. Byrne Page 2 EXPERIENCE (Continued)
' Offshore Loaistics International.-(1977 - i MQ1 Chief Engineer ,
Ensure-safe and. proper operation of vessel and all associated machinery.
Trained and supervised foreign crews (Brazil, Chile). Responsible for maintenance and repair of vessel machinery.
First As'sistant Engineer Assisted chief engineer with vessel's propulsion units.
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1 P . JOHN V..CATALOGNA J .
SR. CONTROLS ENGINEER STONE & MEBSTER ENG'G. CORP.. .
. ACTING MEMBER, BOSTON EDISON CONTROL SYSTEMS GROUP
- : EDUCATION J
? ~B.S.,7 Marine Engineering, Massachusetts Maritime Academy, 1969 LICENSE' LU.S. Coast Guard Third Engineer in Steam and Diesel,' Unlimited
' EXPERIENCE- T
- Stone & Webster Enaineerina Corn. (June 1973 - Present)
Contro1' Systems Engineer, Control Systems.Div., Boston Edison Co. Nuclear
' Engineering Dept. (beginning July.1989).
- Assigned to Boston Edison Co. as seconded engineer to act as' cognizant 7' . engineer for control. panel' improvements for Pilgrim Station Control Room Design Review Project. Preparing drawings and sketches, work
~
-instructions,-. procedure changes- and material-lists; investigating hnd
~ resolving design and. installation problems; co'ordinating with
! construction, operations,.other engineering and consultant personnel regarding implementation details.
.Eastaan' Kodak Het End Rebuild.- Bldg. 319, Sulzer Escher Hyss Inc. (May 1989 - June 1989)
' Responsible' for development of logic diagrams for system operational requirements to be utilized for programming a programmable logic
-controller (PLC). Developed system operation procedures for various modes of. system operations.
Matts Bar Nuclear Power Station - Units,1 and 2, Tennessee Valley Authority (July'1937 - Oct.1987) (Jan.1988 - Mar.1988) (May 1988 - May 1989).
As. responsible, task' engineer, responsible for providing an assessment and detailed evaluation'of a Bypass and Inoperable Status Indication System (BISI) for a-1200 MH pressurized water nuclear power plant to develop the logic for the BISI system for input to the plant's emergency response facilities computer. Further assigned to review and select interface points of all applicable instrument and control systems necessary to support the fluid and electrical systems required to operate Hatts Bar-Unit I while Unit 2 is under construction. The plants utilize common or
, associated systems.
= ' Shortwave II - first Church of Christ, Scientist (FCOC) (Mar 1988 -- May 1988)
Assigned as consultant to FCOC with prime responsibility of reviewing
-contractors documentation provided for the installation and controls of
- the Shortwave II radio transmission station located in Cypress Creek,
-South Carolina.
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- J John V. Catalogna i
- Page 2
' EXPERIENCE (Continued)
Comanche Peak Steam Electric Station. Texas Utilities (Nov.1987 - Dec.
1987)
Assigned to' validate accident monitoring instrumentation with regards to NRC R.G.1.97 post accident monitoring.
Beaver Valley Power Station - Unit 2 Duquense Light Company (Nov.1979 -
June 1987)
- As principal-logic engineer, responsible for scheduling of logic diagrams, preparation of logic diagrams-and system descriptions, review of nuclear steam supplier's functional drawings, review of vendor supplied control systems, control room design review task analysis and emergency operating procedures walk-through, and preparation of FSAR for an 888 MH pressurized water nuclear power plant. Responsibilities also' include providing liaison with client, engineering, and operating staffs.
Millstone Nuclear Power Station - Unit 3, Northeast Utilities Service Company (June 1973 - Oct. 1979)
As control logic engineer, responsible for the preparation of logic diagram and system descriptions for a 1200 MH pressurized water nuclear power plant located in Waterford, CT.
United States Merchant Marine (June 1969 - May 1973)
As engineering officer on merchant ships sailing internationally, responsible for ship operation, maintenance, and repair.
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L
" E.L.-(RETT) CONSIDINE
' CONTROL PANEL DESIGN SPECIALIST MANAGEMENT ANALYSIS CO.
PROFESSIONAL TRAINING I -United. States Navy technica1' schools:
w U.S. Nuclear Power School, Mare Island, California Nuclear Power Training Unit, Idaho Falls, Idaho Electronic Technician A. School, Treasure Island, California Submarine School, New London, Connecticut PROFESSIONAL AFFILIATIONS
-Instrument Society of America American Nuclear Society Human Factors Society Professional Reactors Operators Society EXPERIENCE Manaaement Analysis Comoany (1986 - Present) e Proiect Manager for Boston Edison control room upgrades, performing I analysis _and application of' surface enhancements to the control room panels.. Information from the E0Ps, ops, P& ids, System Descriptions were ;
made part of the control panel through the use of demarcation, labels and meter scales. Additional wcirk entailed the physical rearrangement of components on the control panels. >
Held line management position in the Nuclear Training Department at Rancho Seco Nuclear Generating Station. As superintendent for Administrative Services, was also responsible for the Accreditation efforts. Provided expertise in three~ major areas for assisting control room operator:
control room operations, training, and procedures.
Bechtel Power Corporation. (1969 - 1986)
. Engir,eerinr; . specialist fer control room evaluations r.nd improvements on niajor nuclear power platits.
Staff engineering specialist for South Texas Project in development and implementation of control room design review per NUREG 0700.
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'E.L. Rett Considine "Page 2'
.O EXPERIENCE'(Continued)
Directed the control' systems discipline on a 650 MH coal-fired power
. plant, in:1uding costs, scheduling, procurement, evaluations, budget, and-performance reviews of personnel.
Control systems specialist assigned to a 950 MH nuclear power plant project in Spain. Responsible for development of requirements and preliminary design.for control. room,: computer, and interaction of the control systems on the project.
Control systems supervisor on a seawater pipeline; coordinated implementation of 16.-interactive control rooms, including all analog
. instrumentation and control. logic.
Engineering group leader for the control room design and the control-systems integration of a nuclear steam supply system contract.
Provided proposal and Preliminary Safety Analysis Report technical support for domestic and international efforts.
Served as start-up field liaison for engineering and construction for computer modifications at South California Edison, Alamitos and Huntington Beach generating stations.
Participated on the following projects as staff engineer or in-house consultant to control room design:
San Onofre Units 2 and 3 -- California Lemoniz -- Bilbao, Spain ASCO -- Madrid, Spain A.H. Vogtle -- Georgia Roy S. Nelson Unit 6 (coal-fired) -- Louisiana .
Fayette Power Project. 2 units (coal-fired) -- Texas Sayago -- Bilbao, Spain H.A. Parish 2 Unit (coal-fired) -- Texas Vandellos Nuclear Center Unit 2 -- Madrid, Spain South Texas Project -- Texas United States Navy (1961 - 1969) 1 Qualified Senior Reactor Operator and Chief Reactor Technician.
Supervised reactor operators and technicians, maintenance of reactor control, protection systems and all instrumentation at U.S. Navy nuclear power training unit A1W. Also served as senior reactor control instructor for instrumentation, reactor physics, and reactor control for reactor operator trainees. Member of the Reactor Operator Qualification Board.
Qualified SRO on the USS Shark.
1 B-17
SIBEN DASGUPTA CONTROL SYSTEMS DIVISION MANAGER BOSTON EDISON COMPANY p EDUCATION Electrical Engineer (Power Systems) Northeastern University,1979 H.S. in Engineering Management (Operations Research), Northeastern University, 1973 Master of_ Engineering in Electrical Engineering (Power Systems), Calcutta University, 1969
. Bachelor of Engineering in Electrical Engineering (Power Systems),
Calcutta University, 1967 PROFESSIONAL REGISTRATION Registered Professional Engineer (Massachusetts).
PROFESSIONAL AFFILIATIONS Member of Institute of Electrical and Electronics Engineer Chairman, IEEE Educational Committee, Boston Chapter.
Member of the Working Group of IEEE Nuclear Power Engineering Committee, Section 4.7, Auxiliary Power Systems.
PROFESSIONAL TRAINING Combustion Engineering Nuclear Power Plant Simulator - Training course in Nuclear Power Plant Operation.
Qualification of Safety-Related Equipment for Nuclear Power.
l Generating Stations - Arranged jointly by Drexel University and IEEE.
Kepner-Tregoe Management Training Course.
PROFESSIONAL EXPERIENCE Boston Edison Company ( 1975 to Present)
' Control Systems Division Manager, Nuclear Engineering Dept. (December 1981 to present)
Responsible for all activities within the Division, which includes planning and scheduling, workload assignments, technical assistance and supervision, and developing new standards and work procedures. Duties of the Division consisted of preparation of process and instrumentation diagrams, logical diagrams, schematic diagrams, selection and specification of all instruments and valves, panels layouts and fabrication drawings, loop drawings, tubing and wiring and installation B-18
Siben Dasgupta Page 2 EXPERIENCE (Continued) detail drawings, vendor evaluation, purchasing, field support and startup assistance. Duties also include engineering activities related to Plant Process Computer and Security Comp'Jter Systems.
Senior Electrical Engineer, Nuclear Engineering Dept. (1978 - 1981)
Responsible for review and approval of recommendations for electrical designs prepared by the principal contractors for a new power plant (Architect-Engineer, Nuclear Steam Supplier, Turbine-Generator manufacturer) in the following areas: Station Auxiliary power systems, station auxiliary power system protection, computer applications for load flow and short circuit studies, undervoltage and underfrequency studies, etc. Responsibilities also included design modifications of station auxiliary power systems for an operating plant. This included undervoltage study, relay coordination, electrical equipment selection, equipment qualification, etc.
Instrumentation and Control Engineer, Nuclear Engineering Dept.
(1975 - 1978)
Responsible for logic diagrams, loop drawings, control panel layout ano fabrication drawings, tubing and wiring diagrams as well as installation detail drawings.
Stone & Webster Enaineerina Corp. (1973 - 1975)
Engineer, Control Systems / Advisory Operations Group Served as a startup engineer at Beaver Valley Power Station #1. Resolved acceptance and startup testing problem items (i.e., control logic troubleshooting) in the ficid. Prepared specifications for instrumentation. Prepared bid analyses and recommendations to utility engineers for selection of instrument suppliecs. Designed control loop, logic, panel layout, P&ID and instrumentation installation drawings.
Formulated calibration data for process control loops.
Bril & Howell Communications Co. (1970 -1973)
Engineer, Production Engineering Dept.
Investigated field failures of electronic components wit the aid of computer controlled test system. Investigated manufacturing problems to i determine cause and to recommend corrective action. Set up test methods, trouble-shooting procedures; designed test jigs. Analyzed the production requirements of products and determine the type and sequenze of l operations, establishing work elements, motion patterns and machine cycles. ;
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Siben Dasgupta Page 3 EXPERIENCE (Continued)
Northeastern University (1977 - Present)
- Part-time lecturer in Graduate School of Engineering.
PUBLICATIONS
" Transient Performance of Three-Phase Induction Motors During Sudden Voltage Depressions": Journal of Technology (India) 1969.
" Degraded or Loss of Voltage Protection of Class IE Auxiliary Power Systems in a Nuclear Power Plant"; S. Dasgupta, J. J. Murphy; presented at the IEEE Nuclear Science Symposium, Oct. 1978. Published in the IEEE Nuclear Science Transactions, Feb., 1979.
" Maximum Frequency Decay Rate for Reactor Coolant Pump Motors"; R. S.
Hahn, S. Dasgupta, E. M. Baytch, R. D. Hilloughly; Presented at the IEEE Nuclear Science Symposium, Oct.,1978; published in the IEEE Nuclear Science Transactions, Feb., 1979.
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NORMAN R. EISENHANN SR. INSTRUMENTATION & CONTROL SYSTEMS ENGINEER BOSTON EDISON COMPANY
' EDUCATION 1
B.S., Electronic Engineering Technology, University of Lowell, 1985 A.S., Electronic Engineering Technology, University of Lowell,1982 A.S., New York State Regents External Degree Program, 1979 PROFESSIONAL TRAINING U.S. Navy Training, Nuclear Power School and Prototype Training, 1972.
" Applied Human Factors in Power Plant Design and Operation", General Physics Corporation, 1987 PROFESSIONAL AFFILIATIONS Institute of Electrical and Electronics Engineers.
American Nuclear Society.
Instrument Society of America.
PROFESSIONAL EXPERIENCE Boston Edison Co.. Nuclear Enoineerina Dept. (June 1988 to Present)
Responsible for close-out documentation packages for HEDs and preparing specification for new CR Survey, Inventory, SFTA, etc. Also responsible for supporting the Equipment Qualification Project with updating Equipment Qualification Data Files and reviewing test reports. Responsible for answering Engineering Support Requests submitted to NED by PNPS.
KUSleAIWerav Services (January 1986 - June _RSSl Assigned to the Boston Edison Equipment Qualification Project. Consultant '
to tne Control Systems Group of Boston Edison. Compiled data, reviewed test reports prepared analyre:, and performed calculations using the Arthenius Methodology to complete environmental qualification of control equipment.
Cognizant engineer for Plant Design Changes. These plant modifications included changes to air operators, instrument air systems, ATHS panels, 4160 volt switchgear, and Reactor Protection Systems. Prepared production
' orders to purchase material to support the PDCs. Reviewed calculations for orifice sizing, relief valve sizing and single failure analysis using Boolean Algebra.
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Norman R. Eisenmann Page 2 EXPERIENCE (Continued)
CYGNA Enerav Services (Aoril 1981 - January 1986)
Engineer for the Control Systems Group of Boston Edison. Cognizant engineer for 5 Plant Design Changes (PDC) at Boston Edison. The PDCs included modifications to control panels, local control switches, and shielding of components. Lead Engineer of Cygna on the Boston Edison Pilgrim 79-01B Equipment Qualification team. Duties included equipment specification and test report evaluation to ensure compliance to 00R Guidelines, HUREG-0588, or 10CFR50.49.
Responsible for the preparation of work instructions and procedures for the Equipment Qualification Program. Provided assistance with project budgeting and computerized scheduling.
For the Shoreham Nuclear Power Station Project, prepared new procedures and~ revised existing procedures for processing of vendor technical bullttins, design changes, client interfaces, and administration of clerical workers.
Stone & Webster Enaineerina Coro. (Auaust 1979 - March 1981)
As an Engineering Associate in the Operation Services Division, evaluated equipment for the selection of spare and replacement parts requirements for Consumers Power Company's Midland Station Units 1 and~2. The evaluation encompassed a thorough analysis of vendor information, Final Safety Analysis Reports, specifications and industrial experiences. This effort involved frequent direct contact with equipment suppliers in order to obtain additional data needed to either complete documentation requirements or perform equipment performance evaluations. Duties included defining the parts by interpreting original equipment technical specifications, Quality Assurance packages, equipment qualification requirements and various codes and standards such as ASTM, ASME, A!4ST, and j IEEE.
While assigned to the Engineering Atsurance Division, developed departmental procedures for the Procurement Cor. trol Group.
U.S. Nyvy (1971 - 1979)
Served seven and one-half years on the U.S. Navy Submarine Force as a Nuclear Plant Operater. For three years of this time, assigned as a Nuclear Training Instructor and Leading Electrical Division Petty Officer at the Idaho Nuclear Facility.
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DAVID WILLIAM GERLITS II h SR. SYSTEMS ANALYSIS ENGINEER BOSTON EDISON COMPANY
- EDUCATION B.S.,' General Science: Physics and Chemistry, University of Iowa, 1977.
l1 - LICENSE Senior Reactor Operator, Pilgrim Nuclear Power Station,1986.
p PROFESSIONAL AFFILIATIONS American Nuclear Society Mensa PROFESSIONAL TRAINING-Station Nuclear Engineer (GE BWR) - 1982 Gould SEL Computer Operating Software - 1986 PNPS Simulator Operating Software - CAE Electronics - 1986 Criterion Referenced Instruction and Instructional Materials Development (Hager/ Pipe)'- 1984- j
. PROFESSIONAL EXPERIENCE Boston Edison Company. (1982 to Present)
Sen m Engineer, Systems and Safety Analysis Division Nuclear Engineering Department (1987 - present).
General responsibilities encompass review of plant modifications to ensure Final Safety Analysis Report and regulatory compliance, and review and preparation of safety analyses. Specific assignments include: Lead systems. engineer for the system function and task analysis portion of the detailed control room design review project; lead systems and safety analysis. engineer for the implementation of the nxsdifications resulting from the PNPS safety enhncement program; NRC audit cc-coordinator for inspection of equipment classification, vendor interface, post-maintenance testing, adequacy and reliability of electrical distribution system; individual plant evaluation risk and reliability engineer assisting in the development and review of system descriptions and associated computer models for the PNPS IPE; lead engineer for the implementation of the plant specific. technical guidelines for emergency operating procedures; and lead systems and safety analysis engineer for the PNPS 10CFR50 Appendix R fire protection analysis.
I B-23 1
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L David Hilliam Gerlits II
- Page 2-
, EXPERIENCE (Continued)
Nuclear Training Specialist, Nuclear Training Department (1982 - 1987)
Primary responsibilities focused on preparation and presentation of training material for initial license and licensed operator requalification training programs. Additional-assignments included:
~
project manager for the development of the training material required for INP0' Accreditation of all unlicensed and licensed operators; operations training instructor for simulator software, responsible for detailed review of software model changes required by plant modifications.
United States Navy (1977 - 1982)
Nuclear Trained Division Officer - USS Ulysses S. Grant (SSBN 631)
After completion of Officer Candidate School and Nuclear Power Training, assigned to the ship, and held the billets of Electrical Officer, Reactor Controls Officer, Communications Officer, and Ship's Training Officer.
Managed the 10-14 man divisions responsible for the operation, testing, and repair of ships engineering and communication equipment. Also responsible for scheduling and budget for the training of all ships personnel, and the maintenance of the ship's training records.
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ANTHONY MORSE, JR.
PRINCIPAL CG. TROL ENGINEER STONE & HEBSTER ENGINEERING CORP.
ACTING MEMBER, BOSTON EDISON CONTROL SYSTEMS GROUP EDUCATION Bachelor of Mechanical Engineering, Yale University, 1943 Graduate Courses in Marine Engineering, U.S. Naval Ar.ademy,1943 General Electric Professional Business Management, 195S PROFESSIONAL REGISTRATION Registered Professional Engineer - Massachusetts, Wisconsin PROFESSIONAL AFFILIATIONS Human Factors Society (1982 - 1989)
American Society of Naval Engineers Society of Naval Architects and Marine Engineers PROFESSIONAL EXPERIENCE Stone & Webster Enaineerina Corp.. (October 1973 - Present)
Control Systems Engineer, Control Systems Division, Boston Edison Company, Nuclear Engineer Department (beginning June 1989)
Assigned to Boston Edison Company as seconded engineer to act as cognizant engineer for control panel improvements for Pilgrim Station Control Room Design Review Project. Preparing drawings and sketches, work instructions, and material lists; investigating and recommending design changes; performing human engineering revicWs per NUREG-0700; coordinating with site anc contractor personnel regarding implementation details.
Control Systems Divisicr. Headquarters (November 1982 to June 1989)
As senior control engineer and principal control engir.eer, responsible for various special assignments including an in-house study of the application of cdvanced control concepts and Human Factors Engineering (HFE) principles to nuclear power plants, division representative on a special system transient analysis task force, and the direction of (HFE) activities including presentation and the preparation of propesals, and support to the Comanche Peak and TVA Nuclear Power Plant Projects and the U.S. Army RDX Expansion Project. Also assigned 2s Control Systems Licensing Representative (Sept.1979 to Jene 1989).
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I LAnthonyMorse,'Jr.
Page 2 EXPERIENCE (Continued)-
Modular High Temperature Gas Cooled Reactor (MHTGR) - U.S. Department of l- _ Energy (October 1988 to June 1989)
As human factors ' engineering (HFE) specialist, responsible for the conceptual design of the control room operator's workstation (CROW) and the development of the role of the operator.
Modular High Temperature Gas Cooled Reactor (MHTGR) - Gas Cooled Reactor l Associates (GCRA) (Jan. 1986 - Jan. 1987)
L As lead control engineer, responsible for engineering and design of instrumentation and controls for Stone & Webster's portion of the balance of plant systems and equipment.
Rancho Seco Nuclear Generating Station - 7acramento Municipal Utility District (SMUD) (May 1984 - Dec. 1986)
As. control room design review specialist, assigned to SMUD to their design review team performing the detai. led control room design review (DCRDR) of i: the Rancho Seco Control Room, in accordance with post-TMI NRC requirements.
Duane Arnold Energy Center - Iowa Electric Light and Power Company (Jan.
1984 - April 1985)
As project engineer, responsible for supplying HFE support to IEL&P in the conduct of their DCRDR of the Duane Arnold Control Room.
Kewaunee Nuclear Power Plant - Hisconsin Public Service Corporation (June 1982 - Nov. 1982)
As project engineer, responsible for carrying out the planning phase and producing a program plan report for DCRDR, of the Kewaunee control room.
Control Systems Division Headquarters (Jan. 1981 - June 1982)
As' senior control engineer, special assignments included establishing company positions on appropriate regulatory guides and generic issues established by the NRC. Acting division equipment specialist.
P&bt Beach Nuclear Plant - Units 1 and 2, Wisconsin Electric Power Company (Aug.1980 - Jan.1981)
As assistant project engineer, on project to design and install equipment required to comply with carious post-THI NRC rec.oiren.ents.
Stone & Webster PMR Reference Nuclear Powcr Plant (1974 - 1960)
As lead control engineer, responsible for engineering aM design of instrumentation and controls for the Stone & Hobster PHR Reference Nuclear Power Plant design, utilizing four PWR designs. The work included the preparation of the instrumentation and controls section of the Preliminary Safety Analysis Report; and I&C presentation to the NRC and ACRS, leading
.to preliminary design approvals.
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i Anthony Horse, Jr.
Page 3 EXPERIENCE (Continued)
Assigned similar responsibilities associated with the design of an Interim Spent fuel Storage Facility, and the preparation of a topical report which was approved by the NRC.
Assigned as Coordinator for covering spent fuel disposal studies, and for the performance of study contracts performed by the Reference Nuclear ~'
Power Plant Project for RVA and UNI Nuclear Industries, Richland, Washington.
New Haven - Units 1 and 2, New York State Electric & Gas Corporation . g (1977 - 1979) L As lead control engineer, responsible for the application of the instrumentation and control portion of the Stone & Hebster Reference Nuclear Power Plant design (utilizing Combustion Engineering PWR) to the
- NYSE&G plant requirements and for the preparation of the appropriate sections of the Preliminary Safety Analysis Report. This assignment was ,
in addition to activities associated with the Stone & Hebster Reference Nuclear Power Plant.
Jamesport - Units 1 and 2, Long Island Lighting Company (Oct.1973 - June 1974)
As principal instrument application engineer (March 1974 - June 1974), in the absence of a Lead Control Engineer, responsible for engineering, design, preparation of specifications, bid evaluation, and selection of instrumentation and controls. . . . . _
General Dynamics. Ouincy Shipbuilding Division. (Oct.1967 - Oct.1973)
As senior design engineer, responsible for engineering application, preparation of specifications, bid evaluations, and selection of vendors for automated engine room control systems, main propulsion machinery, and :
components in the fuel oil, lube oil, main feed, main steam, and forced draft air systems, i
PUBLICATIONS ,
1 g Human Factors Apolied in Updated Reference Plant Control Room Design, witn P.E. Knoble and H.E. Hiklund, ANS 1985 Annual Meeting.
Nuclear Power Plant Standardization - Refere:1ce Plart Design, with H.J.L. i Kennedy and B.G. Schultz, ANS 1976 Annual Meeting.
Major Considerations in the Development of a Reference Nuclear Power Plant Design, with H.J.L. Kennedy, Nuclear Engineering International, October 1976.
Blast Freeze for Higher Quality, Baker's Digest, July 1964.
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LEON J. OLIVIER ;
OPERATIONS S CTION MANAGER 1 BOSTON EDIIDN COMPANY l
' EDUCATION j l"
Peterson School of Power Plant Engineering
-Massasoit College
- Foxboro High School, 1967 LICENSES NRC Senior Reactor Operator NRC Reactor Operator-Mass. 2nd Class Engineer-Mass. Assistant Nuclear Power Plant Operator Journeyman Pipe'F1tter PROFESSIONAL EXPERIENCE Boston Edison Co. (197 to Present)
Operations Section Manager (Acting), Pilgrim Nuclear Power Station (May 1989 to Present)
-Responsible for the overall implementation of.all plant operation programs, as well as recommending modifications to related policies and procedures. Acts as offici.a1 'spokesperson for the Section and as an authority in resolving problems of a critical or controversial nature.
This positiori is accountable fcr the efficient and safe operation of plant equipment.
Chief Operating Engineer, Pilgrim Nuclear Power Station (December 1988 -
May 1989)
Responsible for operating, maintain'9g, and refueling the Pilgrim Nuclear Power Station in accordance with the Facility License, Technical Specifications, Operational Quality Assurance Program and Nuclear Organization Policias and. procedures. This position is accountable for the day-to-day operation of plant equipment and operational activities in accordance with approved procedures, including the assignment of operating personnel.
Simulator Divisioef Manager, Training Dept. (1978 - 1988)
Responsible for overall operation, maintenance and modification of the Pilgrim-specific control room simulator located at the Nuclear Organization's Chiltonv111e Training Center. Also responsible for supervision and oversight of staff personnel and additional contractor support, and for simulator capital and expense budgets.
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. Page 2 EXPERIENCEiContinued) b Nuclear Training Specialist -- Simulator Procurement (1984 - 1987)
Responsible for assisting in the design and procurement of. the Pilgrim Station specific simulator. through coordination of internal reviews of design specifications and vendor proposals. This assignment required relocation to the vendor's facility (CAE Electronics, Ltd.) in Montreal, Canada, for a period of nearly three years. Incumbent with these responsibilities was the creation of the Simulator Malfunction Cause and Effect Document, writing and or review of the final Acceptance Test Procedure (ATP), overall responsibility for its implementation, and negotiation of.all data base post-freeze modifications and simulator enhancements.
' Nuclear Watch Engineer, (1981 - 1984) ,
In charge of the plant during assigned shift. Responsible for the safe, l efficient operation of PNPS including unit startup, shutdown, scram recovery, and administrative oversight of yarveillance testing on plant i systems.
Nuclear Operating Supervisor, (1978 - 1981)
Performed diversity of tasks focusing on the supervision for the operation of the-control room facility in eccordance with station guidelines, Responsible for maintaining awareness of station conditions, supervising the Nuclear Plant Operator and implementing operating maneuvers in accordance with policies and procedures. Assisted in training Nuclear Plant Operators in tasks required for operation of control facilities.
Developed and implemented log and record system of plant operating data.
Nuclear Plant Operator, (1975 - 1979)
Interfaced with supervisors and PNPS operating groups for completion of assigned tasks in the maintenance of overall station equipment.
Identified and reported items requiring specialized attention. Responsible for performing lubrication checks of station equipment. In charge of shutting the reactor down when determined the safety of reactor is in jeopardy or when operating parameters exceed any of the reactor protection system setpoints and automatic shutdown does occur.
Central Control Operator, Mystic Station Responsible for startup and shutdown and operation of hign pressure forced flow C-E boilers and station auxiliary equipment. Performed startup and shutdown of GE high pressure tandem compound Turbine-Generator and accompanying auxiliary equipment.
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r ~, EDWARD L. QUINN SR. CONSULTANT (I&C)-
. MDM ENGINEERING CORP.
- (SUBCONTRACTED TO PICKARD, LOHE & GARRICK)
E
. EDUCATION
'M.S., Management, Rensselaer Polytechnic Institute, 1980.
B.S., Electrical Engineering, Tu.fts University,1976.:
PROFESSIONAL AFFILIATIONS
. American Nuclear Society Society for Task Analysis Institute of: Electrical and Electronics Engineers, Electromagnetic Interference Society Instrument Society of America-
' Toastmasters, International PROFESSIONAL TRAINING Massachusetts Instit'ute of Technology Reactor Safety Course, July 1985 PROFESSIONAL EXPERIENCE MDM Enaineerina Corp. (1988 - Present)
Senior Training Specialist, San Onofre Nuclear Generating Station (May 1988 - Present)
Responsible for developing and revising lesson plans to meet INPO requirements for licensed and'non-licensed operators for SONGS units 2/3.
Senior Consultant, Bostan Edison Company, Control Room Design Review Project (1987 - Present)
Responsible for preparation of utility's procedures for evahtation and categorization of identified Human Engineering Discrepancies, 'ising techniques of reliability analysis and risk assessment; and for procedure to document, track and close-out discrepancies in accordance with NUREG-0700.
Plckard. Lowe and Garrick. Inc. (1986 - 1988)
' Senior Consultant in the areas of instrumentation and control and electric power distribution to support utility maintenance and plant betterment programs, probabilistic risk analysis (PRA), and availability and reliability studies conducted by PLG. Initial studies include control room design review, a setpoint calculation evaluation, development of an B-30
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- Edward L. Quinn Page 2 EXPERIENCE (Continued) l 1
. ALARA (as low as reasonably achievable) Design Manual, and availability
. analyses.for anticipated transient without scram station blackout and-power-operated relief valve. reliability in support of PRAs. Developed an instrument setpoint control program for a utility and a methodology for calculating instrument setpoints based on ISA Standard S67.04, Revision
- 1. Also served as principal investigator on a risk-based surveillance test interval extension program for a nuclear instrumentation system; provided reliability engineering support-to a control room design review program developing methods.for prioritizing the risk of Human Engineering Deficiencies.
Start Un Nuclear. Inc. (1984 - 1986)
Vice President of 65-person consulting' firm and site leader for 40 contract and consulting employees providing varied engineering services at San Onofte Nuclear Generating Station.
Southern California' Edison Company. (1982 - 1986)
San Onofre Nuclear Generating Station (SONGS)
Environmental qualification coordinator. Responsible for implementing all environmental qualification-related design changes on San Onofre Unit 1 (Westinghouse pressurized water reactor) required before startup in 1986:
design review, scheduling, project management, startup testing, and turnover for 60 design packages, accounting for more than 501. of the critical path outage time.
Data' Development Engineer Instrumentation and control and electrical support for Nuclear Information
- Systems- Department under contract with Start Up Nuclear Inc.
Instrumentation and Control Startup Engineer
-Lead project engineer for all accident monitoring systems (AMS) on SONGS 2 and 3,' including installation of the critical function monitoring system
.(CFMS) and the qualified safety parameter display system (QSPDS).
Responsible for design review, construction management, startup testing, and surveillance issue for the HSPDS system and program management for the CFMS installation. Conducted two NRC audite and tours of the AMS installation at San Onofre.
Power Ascension Engineer, SONGS 3 Responsible for power ascension on-shift testing.
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-Edward L.;Quinn ~
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EXPERIENCE-(Continued)
Pennsylvania Power and Light Company (PP&L)'. 1980 - 1982 Electrical Design Engineer, Electrical Separation Task Force. Susquehanna Steam Electric Station (SSES).
. Directed the electrical separation review for Unit:1 for conformance with licensing commitments. Supervised onsite review team composed of Bechtel, General Electric, and PP&L engineers conducting panel and raceway
. inspections of all Class lE equipment for separation violations, qualifying high temperature materials for application as inner-panel barrier, and reviewing design with the NRC. Issued the current utility guideline for all electrical separation used at SSES.
Electromagnetic Interference Study, SSES 1, for effect of portable radios on critical plant instrumentation. Administered PP&L contract with Don White Associates for test program and national laboratory analysis of sample plant components.
Three Mile Island (TMI), GPU Nuclear Corporation and Metropolitan Edison Company. ALARA;(as low as reasonably achievable) Review Engineer.
Reviewed procedures and generated design drawings and flush procedure to separate the common primary sample piping and sink-of Units 1 and 2.
General Dynamics Corporation. Electric Boat Division. (1976 - 1980)
Assistant Project Manager, SSN 702-USS Phoenix. Directly supervised six engineers on three shift rotations who conducted all primary plant testing with Navy crew on board.
Alpha Sea Trials Director, SSN 697-USS Indianapolis. Responsible for primary plant. initial full-speed surfaced and submerged runs and a team of eight data takers.
Power Ascension Director, SSN 697-USS Indianapolis. Responsible for all initial cr'ticality and power ascension activities on board and at the power range test center.
Shift Test Engineer. Supervised reactor plant construction and directed j testing from initial turnover to final acceptance.
PUBLICATIONS Quinn, E. L., " Risk-Based Evaluation of Surveillance Test Procedures at San Onofre Nuclear Generating Station," accepted for presentation at the American Nuclear Society Summer Meeting, San Diego, California, June 13, 1988.
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Edward L. Quinn
.Page 4 EXPERIENCE (Continued)
Quinn, E. L., " Baseline Characterization of Electrical Circuits To Support Plant Life Extension," presented at the American Nuclear Society Winter Heeting, Washington, D.C., November 20, 1986.
Quinn, E. L., " Prototype for an Automated Technical Specification Information System," presented at the American Nuclear Society Hinter Meeting, San Francisco, California, November 14, 1985.
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E KENNETH NORNAN TAYLOR NUCLEAR NATCH ENGINEER (OPERATIONS SECTION STAFF)
OPERATIONS LIAISON TO DCROR EDUCATION Currently attending Northeastern University pursuing a degree in engineering.
Nuclear Power Training Unit, West Milton, NY - 1960 U.S. Navy Power School, New London, CT (1959)
Machinist's Mate "A". School, Great Lakes, IL Cole Trade High School, Southbridge, MA LICENSES Mass. Nuclear Power Plant Operating Engineer (1978)
NRC Senior Reactor Operator License (1977)
NRC Reactor Operator (1975)
Mass. License - 1st Fireman (1975)
PROFESSIONAL EXPERIENCE Boston Edison Company Operationr, Section - Staff SRO (1988 - Present)
Responsible for various staff assignments related to nuclear operations, including staff liaison between DCRDR Project and Operations Section.
Reviews prospective design changes, coordinates operation review and comment, and helps to ensure coordination between Operations and Engineering on major project activities.
Day Natch Engineer-Pilgrim Nuclear Power Station (2/81 - 1982)
Responsible for the safe, efficient operation of Pilgrim Station, under the direction of the Chief Operating Engineer in accordance with the squirements of Station Procedures and Regulatory Agencies. Responsible for rewriting procedures, update of P&ID's and ensuring a smooth accurate communication with the departments within the station.
Nuclear Match Engineer-Pilgrim Nuclear Pcwer Station (11/78 - 2/81 and 1982 - 1988)
Responsible for all activities relating to station and safety including, fuel loading, startup and shutdown in accordance with the requirements of the operating license, Technical Specifications, approved operating procedures, reg',latory agencies, and the Operations Quality Assurance Program. Respe nsible for implementing the station radiation protection program, for t.ie monitoring the performance of station equipment, for assuring that the reactor is shutdown when a condition has been identified such that continued operation would jeopardize station safety and the station security within the confines of the process building.
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Kenneth N. Taylor
!Page 2-EXPERIENCE'(Continued)
Nuclear' Operating Supervisor-Pilgrim Nuclear Power Station (11/75 -'11/78)~
Responsible for supervising.the Nuclear Plant Operations and implementing operating' maneuvers in accordance with approved station' procedures and.for.
. assisting.in training the Nuclear' Plant Operators in the-skill and knowledge required for.the safe and efficient operation of a nuclear facility.
'U.S. Navy.
5/73 to 11/75. . .
Served on USS Skipjack SS(N) 575 as Engineering Hatch Supervisor 4/72 - 5/73 Served on staff at Engineering Repair Division, New' London Conn.
8/65 - 4.72 Served on USS Francis Scott Key SSB(N) 657 as Engineering Officer of the Hatch 12/62 - 8/65 Served on USS Stonewall Jackson (SSB(N) 634 as Engineering Hatch Supervisor 1/61 - 12/62 Served on U.S.S. Ethan Allen SSB(N)607 as Engine Room Supervisor 1/59 - 1/61
. Received US Naval Training at various schools 2/57 - 1/59 Served on USS Skate SS(N) 578 as Engineer Room Operator 12/56 - 2/57 Served on USS Leyte C.V.S 32 as Auxiliary Operator.
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KATHLEEN D. HARD SR. SYSTEMS AND SAFETY ANALYSIS ENGINEER BOSTON EDISON COMPANY EDUCATION H.S., Radiological Sciences, University of Lowel!, 1988 B.S., Radiological Sciences, Lowell Technological Institute,1975 Additional graduate courses, Business Administration (Computer Science),
Florida Institute of Technology (1979 - 1982)
LICENSE Reactor Operator, St. Lucie I Nuclear Power Station, 1980 (expired).
PROFESSIONAL AFFILIATIONS American Nuclear Society PROFESSIONAL TRAINING
" Applied Human Factors in Power Plant Design and Opeation," General Physics Corp., 1987.
PROFESSIONAL EXPERIENCE Boston Edison Company (October.1985 to Present)
Systems and Safety Analysis Division, Nuclear Engineering Department Project Manager for the Individual Plant Examination. Involved in special projects including the Availability Improvement Program, Detailed Control Room Design Review, root cause investigations, and the coordination of NRC it.spections. Responsible for maintaining the safety and power generation design basis of Pilgrim Station through the inter-system review of design changes and safety evaluations.
Florida Power and Licht Co.
Nuclear Reactor Operator, St. Lucie 1 Nuclear Power Station (1980 - 1982)
Responsible for the operation and safety, as well as associated documentation, of a 2560 MHt Combustion Engineering Pressurized Hater Reactor. Involved in plant evolutions including shutdown, cooldown to refueling mode, new and spent fuel shuffle, and reactor startup and heatup to full power operation. Selected to be the Reactor Operator for the full scale emergency exercise.
Senior Plant Technician. St. Lucie 1 Nuclear Power Station, (March 1979 -
February 1980) i Responsible for operatire, refining, and documenting the computerized j Health Physics records ma.xgement system. Pioneered an effort to develop specifications for an on-lir.e, real-time computerized records management system. Trained plant personnel on exposure limits and ALARA dose control.
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' EXPERIENCE (Continued)
' Yankee' Atomic Electric Co.
Radiation Physicist, Environmental Engineering' Department Project Manager of the Thermoluminescent Dosimetry (TLD) Section of the Radiation Protection _ Group. Responsible for the design, development, and-implementation of a personnel dosimetry program known as the Total Radiation Assessment Program (TRAP) including TLD hardware and its
. interface with supportive software. As part of the Environmental Laboratory Group, supervised the operation and calibration of radiochemical environmental sample counting equipment. Coordinated a special effort to develop and implement an in 111u spectrometry system. 1 PUBLICATIONS
" Computerized Radiation Exposure Records Management," R.D. Schauss, K.D.
Hard, and-R.J. Freddie Presented at the 22nd Annual Health Physics Society Meeting; July, 1977
" Applications of a General Equation.for the Kinetics of Linear First.Oro~er Phenomena," K.H. Skrable, C. French, G. Chabot, and K.D. Hard Nuclear Safety' Journal; April, 1975 l
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