ML20236N259

Final Summary Rept of Human Factors Engineering Review for Byron & Braidwood Stations Spds

ML20236N259
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	07/31/1987
From:	COMMONWEALTH EDISON CO.
To:
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ML20236N234	List: ML20236N231 ML20236N259
References
NUDOCS 8708110426
Download: ML20236N259 (54)

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COMMONWEALTH EDISON COMPANY FINAL

SUMMARY

REPORT OF THE HUMAN FACTORS ENGINEERING REVIEW FOR THE BYRON AND BRAIDWOOD STATIONS 1 SAFETY PARAMETER DISPLAY SYSTEM JULY, 1987

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1 8708110426 070730 PDR ADOCK 05000454 P PDR c.

TABLE OF CONTENTS SECTION PAGE

1. 0 INTRODUCTION 1-1

2. 0 OVERVIEW 2-1 2.1 Data Collection Phase 2-2 2.2 Findings Assessment Phase 2-2 2.3 Reporting Phase 2-2

3. 0 MANAGEMENT AND STAFFING 3-1

4. 0 DOCUMENTATION AND DOCUMENT CONTROL 4-1 4.1 Input Documentation 4-1 4.2 Output Documentation 4-2

5. 0 INTEGRATION WITH OTHER SUPPLEMENT 1, NUREG-0737 INITIATIVES 5-1

6. 0 REVIEW PROCESSES 6-1 6.1 Checklist Survey 6-1 6.2 Personnel Survey 6-2 6.3 Evaluation of Parameters Displayed on the SPDS with Respect to the DCRDR Task Analysis /

Validation Data 6-3

7. 0 FINDINGS ASSESSMENT 7-1

8. 0 IMPLEMENTATION OF CORRECTIVE ACTIONS 8-1

9. 0 REFERENCES 9-1 APPENDICES A SPDS QUESTIONNAIRE USED FOR THE BYRON AND BRAIDWOOD STATIONS CONTROL ROOM PERSONNEL INTERVIEW A-1 B PERSONNEL DEMOGRAPHICS OF BYRON AND BRAIDWOOD STATIONS SPDS INTERVIEW PARTICIPANTS B-1 C FINDINGS IDENTIFIED BY THE BYRON AND BRAIDWOOD STATIONS SPDS PERSONNEL INTERVIEW C-1 D FINDINGS IDENTIFIED BY THE BYRON AND BRAIDWOOD STATIONS SPDS CHECKLIST REVIEW D-1 E FINDINGS IDENTIFIED BY THE BYRON AND BRAIDWOOD STATIONS SPDS/DCRDR TASK ANALYSIS REVIEW E-1

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1.0 INTRODUCTION

Among the directives issued to the nuclear power industry by the Nuclear Regulatory Commission (NRC) in the aftermath of the accident at Three Mile Island Unit 2, was the need to develop a Safety Parameter Display Sytem (SPDS) for each generating station. The purpose of the SPDS is to display in a single location the value/ status of primary variables which directly indicate the status of the safety parameters which indicate the accomplishment or maintenance of plant safoty functions. The display should function to aid the control room personnel during abnormal emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective actions by the operators to avoid a degraded core.

Following the issuance of HUREGs 0696 (Reference 3) and 0737 (Reference 4) in 1980, Commonwealth Edison Company (CECO) began the design of the SPDS for its operating stations and stations under construction. Although human factors were taken into account during the initial design of the systems, CECO committed, as part of its April 14, 1983 response (Reference 1) to NUREG-0737, Supplement 1, to a human factors review of the SPDS at each of its nuclear stations. The purpose of this review was to ensure that the design of the installed SPDS complies with sound human factors engineering principles and to integrate the SPDS work with the human factors efforts associated with the other Supplement 1 initiatives.

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The present SPDS human factors review is also seen by CECO as an integral part.of the SPDS Verification and Validation I process.

An important aspect of the review was an evaluation of whether or not the plant variables displayed on the SPDS are sufficient for the operators to assess the safety status of the plant. This evaluation was conducted with reference to the task analysis data collected during the Detailed Control Room Design-Review (DCRDR) at Byron and Braidwood Stations.

These data, in conjunction with the data collected during the administration of those sections of the SPDS Checklist Survey and the SPDS Personnel Survey (see Section 2.1) that address parameter selection, were used to verify the need for

- thome parameters that are currently presented by the SPDS as well as identify the need for any additional parameters.

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I 2.0 OVERVIEW The human factors review of the SPDS evaluated the extent to which the display provides the necessary information for the control room operator to determine the safety status of the plant. The SPDS was evaluated in terms of the following; o Appropriateness and completeness of the information available through the system ( i. e. , does the SPDS provide the parameters and variables necessary to determine the status of the critical safety functions of the plant?)

o Effectiveness of the display format and coding techniques that are used to draw the operator's attention to important information o Location and position of the SPDS CRTs in the control room o Readability of the information given the display hardware and environmental factors such as lighting ,

and glare o Adequacy of procedures and documentation for interpreting the display j The SPDS review was conducted as a three-phase process. The first phase consisted of several data collection activities j that provided the basic data from which human factors problems were documented. The second phase consisted of an assessment of the findings identified in the data collection phase. The third phase consisted of reporting the results of the review. The present report is the product of this third phase.

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2.1 Data Collection Phase There were several activities involved in the review that provided data for. consideration:

I o A Human Factors Checklist Survey of the SPDS and j its implementation in the Byron and Braidwood control rooms o A Personnel Survey consisting of structured interviews of operations personnel at various levels o An evaluation of the variables and parameters selected for inclusion in the SPDS display with reference to the Task Analysis data collected during the Byron and Braidwood DCRDR's A more detailed description of the methods employed in each activity and the findings that resulted are presented in Section 6.

2. 2 Findinas Assessment Phase An assessment team was formed to evaluate the Human Engineering Findings (HEFs) identified during the Data Collection Phase. This assessment team included representatives from the plants, general office, and human factors personnel. The team recommended a resolution for each finding. -The corrective action to be taken or a Justification describing why no modification is necessary was determined by the team.

2. 3 Reportina Phase The present report represents the methodology, findings and conclusions from the Byron and Braidwood Stations SPDS review. The Byron and Braidwood SPDS Findings are presented in Appendices C, D, and E. This report was prepared to show compliance with CECO's April 14, 1983, commitments to the NRC (Reference 1).

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3.0 MANAGEMENT AND STAFFING The human factors review of the Byron and Braidwoods SPDS was I conducted through the cooperative efforts of CECO and ARD (Advanced Resource Development) personnel. The review team met or exceeded the CECO commitments in Reference 1 and included well qualified and experienced personnel in the areas of nuclear operations, engineering, human factors I engineering, and computer systems. Both Ceco station and general office personnel participated.

The CECO effort was headed by the Human Factors Engineering Coordinator and Alternate Human Factors Engineering Coordinator (R.E. Howard and K.A. Hesse) in the Nuclear Services Technical (NST) Department. They have participated in human factors activities at each of CECO's nuclear power stations and had coordinated the DCRDR and Emergency Response Facilities (ERFs)

Pro]ects at the Byron and Braidwood Stations. They were assisted by upper-level plant personnel and personnel from the Station Nuclear Engineering Department (SNED).

Personnel from both Byron and Braidwood Stations served as Subject Matter Experts (SMEs) and supported the Human Factors Engineering Team, as needed. The SMEs included personnel familiar with the computer systems, operating procedures, and instrumentation in the control room.

The Lead Human Factors Engineer (R.L. Kershner, ARD) was supported by senior and staff-level human factors engineers with appropriate experience in nuclear industry human factors. In order to promote the integration of the SPDS review with other 0737 initiatives, a number of the human 3-1

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l factors personnel who supported the Byron and'Braidwood SPDS  !

review' vere selected from those who had previously supported the Byron and Braidwood Stations DCRDR. The human factors I

personnel were' involved in the human factors review of the

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Byron and Braidwood Emergency Response Facilities (ERFS) 1 during the same time frame as the SPDS review.

The Byron and Braidwood Stations assessment team consisted of senior staff and plant management personnel whose back-grounds and experience included aspects of:

1. . Plant Operations

2. Station Nuclear Engineering

3. Human Factors Engineering

4. Computer Systems i

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4. 0 DOCUMENTATION AND DOCUMENT CONTROL  !

' 4,1. Input Documentation The review team used the following documents to support the review processa o Drawings of the physical layout of the Byron /Braidwood Control Rooms o Human Factors Checklist developed by CECO for use in the DCRDRs o Design specifications for the SPDS and Prime Computer displays o DCRDR task analysis data and instrumentation requirements list o The CECO Supplement i submittal letter to the NRC (Reference 1) o NUREGs 0696 (Reference 3), 0737 (Reference 4),

0835 (Reference 5), 0800 (Reference 6) and 0700 (Reference 7) o CECO SPDS Requirements Document (Reference 8) 1 4-1

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4. 2 Output Documentation L

In addition to the present report, the following documents were generated during the review process:

o SPDS Checklist derived from input documents #2, 5,'and 7 specific for CECO SPDSs o SPDS Checklist cross-referenced to CECO SPDS Requirements Document o Completed SPDS checklist for the Byron and Braidwood SPDS o Transcription of responses to the SPDS Personnel Survey o Summary of content analysis or responses to the SPDS Personnel Survey o . Copy of findings as initially presented to CECO by ARD 4-2

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5. 0 INTEGRATION WITH OTHER SUPPLEMENT 1, NUREG-0737 INITIATIVES Commonwealth Edison Company has an integrated program to address each of the Supplement i to NUREG-0737 initiatives.

This program extends throughout its system of nuclear generating stations and has specific provisions for each station. This program provides the necessary integration and support to ensure that a systematic approach is adopted for the inclusion of each of the recommended design changes resulting from these initiatives. Details of this process, including schedules were provided in Commonwealth Edison's April 14, 1983 submittal to the NRC (Reference 1).

At each station, the design of the Safety Parameter Display System (SPDS), the Regulatory Guide 1.97-based instrument displays, the development of function-oriented emergency operating procedures, the training of the operating staff, the DCRDR, and the ERF reviews are being integrated in a l manner which takes full advantage of the scheduling of each l of these initiatives. The human factors engineering review of the SPDSs is being conducted after the DCRDR at each station and after the operational date for the SPDS. By performing the SPDS review after the DCRDR, it is possible to better integrate the data collected and the findings derived from these two activites. By performing the review after the operation date of the SPDS, it is possible to obtain more meaningful input from operations, because by this time, they have had experience with the system.

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The SPDS review for the Byron and Braidwood Stations was also integrated with the human factors engineering review of the Byron /Braidwood ERFs, i. e. , the Technical Support Center (TSC) and Emergency Operating Facility (EOF), which overlapped in time with the SPDS review. The SPDS is available in.the CECO ERFs, as well as in the control room at each station, and as such, it provides information to TSC/ EOF personnel. Findings related specifically to the use of the SPDS displays in the ERFs will be documented in the report on the ERF reviews. Findings related to the use of the SPDS in the control room, or more generally to the use of the SPDS at all sites where it is available, will be documented in the present report.

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6. 0 REVIEW PROCESSES 6.1 Checklist Survey A checklist survey was conducted to determine whether the information displayed by the SPDS, and the equipment and procedures used to access the system, conform to sound human factors engineering design principles. Features under invest-igation included the display of critical plant variables, the use of color as an indicator of safety status, CRT specifica-tions, and equipment placement and accessibility. The checklist was applied across all modes of plant operation.

The checklist was developed in the following manner. In 1982, CECO developed a preliminary human factors checklist for their SPDS based on NUREG-0700 (Reference 7) and NUREG-0835 (Reference 5). This checklist was used to review the SPDS prior to its implementation at the stations. During the Dresden Station SPDS review, this 1982 checklist was updated based on relevant. sections from the recently issued appendix to NUREG-0800 (which constitutes the final version of NUREG-0835) and the checklist derived from NUREG-0700 that CECO developed to support the DCRDRs. The updated checklist (1985) encompassed all issues that had been addressed by the 1982 checklist.

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e A copy of the resulting SPDS checklist was completed by human factors engineers with assistance, as needed, by CECO SMEs. Items of non-compliance to the SPDS checklist were noted as HEFs. These HEFs are included in Appendix D.

The CECO Human Factors DCRDR checklist was also examined to determine if there were any findings from the DCRDR that were also applicable to the SPDS. These findings vere assessed during the conduct of the Byron /Braidwood DCRDR and are included in the DCRDR Final Summary Report submitted to the NRC (See Reference 9).

6. 2 Personnel Survey Structured interviews were conducted by human factors engineers on a one-to-one basis with licensed operating I personnel for the Byron and Braidwood control rooms. The questionnaire presented in Appendix A was used as a basis for the interviews. The interview, which lasted approximately one hour with each of twelve individuals, was structured to provide information regarding the following areas:

o Perceived usefulness of the SPDS o Location of SPDS CRTs in the Control Room o Display characteristics o Use of coding techniques, labels, color o Integration with annunciator alarm system o Procedures and training for use of SPDS o Availability of documentation Interviews were conducted with a representative sample of Nuclear Station Operators (NS0s) and Shift Control Room Engineers (SCREs) from Byron and Braidwood Stations. The interviewer verbally asked the questions of each participant and noted the response. Care was taken at all stages to protect the confidentiality of the participants' responses.

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Notes on the interview' responses were transcribed and responses from all participants were. compiled on a' question-by-question. basis. A content analysis was performed and a t.

listing of issues was compiled.- Frequency counts of the number'of' participants from a given position (NSO, SCRE, SF, SE) 'who had mentioned each issue'were obtained. Each issue was then categorized by human factors personnel as:

o A human factors engineering problem o' A correct comment not considered a human factors engineering preblem o An incorrect comment; this was based on inconsistencies.with other comments and with additional ~1nformation'available to the review team from various sources o A general comment or opinion

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6. 3 - Evaluation of Parameters Displayed on the SPDS with Respect to the DCRDR Task Analysis / Validation Data During the conduct of the Byron /Braidwood DCRDRs a j

comprehensive task analysis was per' .se d . The purpose of the task analysis was to identify -

controls and displays necessary in the control room to y cmit operators to effectively mitigate emergency symptoms. The basis of the analysis was the! Pressurized Water Reactor (PWR)

Westinghouse Owners Group Emergency Response Guidelines (ERGS), Rev. 1 (Reference 8). -The review was conducted to partially fullfill regulatory requirements as set forth in Supplement 1 to NUREG-0737. Because the DCRDR task analysis was designed to be comprehensive and to identify the displays used to mitigate critical emergency conditions, the results provide a medium for verification of the SPDS parameter selection, as well as a vehicle to integrate the SPDS with other Supplement 1 initiatives.

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i To assure:that the SPDS adequately reflects'the parameters that are. critical-for monitoring plant safety-status during emergency conditions, as well as in pre- and post-accident conditions, a comparison and evaluation was conducted.between the displays and indicators determined to~be relevant to mitigating emergency. conditions during the station's DCRDR

.and the variabies displayed on the SPDS. The approach described'belcw was used to implement the comparison and evaluation.

The data collected in the DCRDR task analysis were entered into a computerized database management system. This database was-manipulated to produce a listing that contained all the

' displays, including indicator lights, identified in the task ,

analysis.- Specifically, the listing contained the following information for each displays o Equipment identification number o- Parameter measured o Units of' measurement

.o Task number (cross-indexed back to the WOG ERGS) in which the display was identified In addition, for each task number presented, the listing contained the required range for the parameter being displayed as well as the divisional' increments in which that range should appear. This list was compared to the SPDS parameters by a Human Factors Specialist (HFS) who was familiar with the station's task analysis data, working with a station SME.

The comparison and evaluation proceeded in the following steps:

1. Determine which task analysis dispirys and indicators are currently being displayed on the SPDS or serve as input to the calculations that i'

produce the variables that are displayed. Confirm that each of these displays and indicators reflects the status of a safety parameter which indicates the accomplishment or maintenance of a plant safety p function.

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2. Determine those task analysis displays and

-indicators that do not fall within the. scope of

' the SPDS in that they do no'c indicate the status of safety parameters which indicate the accomplishment er. maintenance of plant safety functions.

3. ' Determine those' task analysis displays and indicators that may fall within.the intended scope of'the SPDS but which are not currently reflected in any SPDS input or variables.

4. Determine whether the SPDS contains variables or input'that are not substantiated by displays or indicators identified in the task analysis.

5. Determine whether or not the variables represented in the SPDS reflect the required range as identified in the task analysis.

The comparison and. evaluation was conducted from the perspective of operating requirements, that is, what should be as. opposee to what currently exists. Limitations of the

.present har 1-re, computer software and/or display formats were not c oidered during the evaluation.

The findings of this task analysis confirmed that the variables currently displayed on the SPDS reflect the status of parameters which indicate the accomplishment or maintenance of plant safety functions. However, additional parameters ar.d/or displays were identified as desirable to have during transient events and were documented as Human.

Engineering Findings (HEFs). Three HEFs regarding the SPDS that were, generated.as a result of the task analysis review are contajtned in Appendix E.

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7. 0 FINDINGS ASSESSMENT

' The Human Engineering Findings (HEFs) compiled from the data '

collection activities, presented in Appendices C,-D and E included the following information:

1. Finding number

.2. Checklist guideline number that the finding violated, if applicable

3. Data collection activity from which the finding resulted

4. Description of the problem

5. Resolution of the finding The Byron and Braidwood Stations assessment team collectively arrived at a resolution for each finding. As the result of clarifications or additional information that arose in the course of the assessment process, it became apparent that some findings were not valid, either because of a misinterpretation of the data from which the finding had been written or because the problem specified in the finding had already been corrected.

These invalid findings were then cancelled.

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8. 0 IMPLEMENTATION OF CORRECTIVE ACTIONS y

The ultimate responsibility for addressing the HEFs discovered in the SPDS review process rests with the Station .

Operations and Station Nuclear Engineering Departments. The assessment team's recommendations with. regard to the resolution of the f1ndings were reviewed by appropriate representatives of these departments, in con] unction 1with human. factors personnel, and final decisions were made as to which findings warranted correction.. Justifications were written for those findings'that warranted no further. action.

These, Justifications are presented in Appendices C, D and E.

For each finding for which.a corrective action was agreed-upon, a CECO staff member with appropriate responsibility was identified. This staff member then determined a time frame for the implementation of the corrective action. These implementation schedules are also documented in the applicable appendices.

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9.0 REFERENCES

1. April 14, 1983 letter from Cordell Reed, Commonwealth Edison,. to Harold Denton, U. S. Nuclear Regulatory Commission, regarding CECO's response to NUREG-0737, Supplemental 1.

2. U. S. Nuclear Regulatory Commission, " Clarification of TMI Action Plan Requirements," USNRC Report NUREG-0737, Supplement 1 (Generic Letter 82-33), 1982.

3. U. S. Nuclear Regulatory Commission, " Functional Criteria for Emergency Response Facilities," USNRC Report NUREG-0696, February, 1981.

4. U. S. Nuclear Regulatory. Commission, " Clarification of TMI Action Plan' Requirements," USNRC Report NUREG-0737, November,~1980.

5. U. S. Nuclear Regulatory Commission, ." Human Factors Acceptance' Criteria for the Safety Parameter Display System," USNRC Report NUREG-0835, October, 1981.

6. U. S. Nuclear Regulatory Commission, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," USNRC Report NUREG-0800,. Revision O of Appendix A to SRP Section 18.2,

" Human Factors Review Guidelines for the Safety Parameter Display System (SPDS)," January, 1985.

7. U. S. Nuclear Regulatory Commission, " Guidelines for Control Room Design Reviews," USNRC Report NUREG-0700, September, 1981.

8. Marianyi, J. J., Wong, B. M. K., Tate, L., & Dages,'K. M.

" Commonwealth Edison Company Safety Parameter Display

. System Requirements Document -- Revision 2," January, 1985.

9. Commonwealth Edison Company, " Byron /Braidwood Stations Detailed Control Room Design Review Final Summary Report",

December 1, 1986.

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p 4 i APPEf!IDX A

-SPDS QUESTIONNAIRE USED FOR THE BYRON AND BRAIDWOOD STATIONS CONTROL ROOM PERSONNEL-' INTERVIEW l

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SPDS INTERVIEWS -- COVER LETTER ARD Corporation is supporting Commonwealth Edison Company in a human factors review of the CECO Safety Parameter Display System (SPDS) and other CRT displays that are available in the Control Room at each station. The SPDS review will support the SPDS Verification and Validation Program from a human factors perspective. This particular segment of the review pertains specifically to the SPDS and CRT displays at Byron (or Braidwood) Station. As part of this effort, we are  !

interviewing control room personnel who have had working experience with the SPDS and other displays. Our goal is to determines (1) what information should be available on the SPDS, and 1 A

(2) how this information should b4 displayed (i.e.,

what grouping,-format, user-system interface).

The interviews will address the issues listed in the attached questionnaire. THERE IS NO NEED FOR YOU TO WRITE RESPONSES TO THESE QUESTIONS. We are distributing the questionnaire now so that, if your schedule 'ermits, p you can familiarize yourself with the issues that will be raised during the interviews.

Your input will be valuable to ur in documenting what the Operations staff views to be the strong points of the present SPDS, in identifying potential human factors problems related to this display, and in formulating recommendations that will both correct any identified problems and respond to recent Nuclear Regulatory Commission requirements.

We anticipate that each interview will last about one hour.

The interviewer will be making notes on your verbal responses to the items on the questionnaire. These notes will be transcribed and then combined with those of your colleagues and summarized on a question-by-question basis.

Your responses will be strictly confidential. Only ARD personnel will have access to the transcribed notes, and the l

findings presented to CECO management will be based on the summaries. The demographic information that we request from you sill be dissociated from your responses to the questionnaire.

1 Any potential human factors problems will be documented in the form of Human Engineering Findings (HEFs). These HEFs will then be assessed and resolved by an HEF Assessment Team consisting of CECO managers and a representative from ARD.

ARD will also support CECO in preparing a final report to the NRC.

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L SPDS INTERVIEWS --' DEMOGRAPHICS

1. Present' position:

How long?

2. How long with Commonwealth Edison?

3. How long in. nuclear industry?

4. Previous industry positions:

5. Do you hold a Reactor Operator (RO) license? ,

How long?

Which station (s)?

6. Do you hold a Senior Reactor Operator (SRO) license?

How long?

Which station (s)?

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- Date Interviewer Station Interviewee #

SPDS (TOP LEVEL OVERVIEW OF PLANT STATUS)

- General Information

1) How long have you used the SPDS?

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! 2) Have there been any transients or abnormal events since I you have been using the SPDS? If so, did the display.  !

prove to be effective in monitoring plant status during this event? If not, why?

3) Has the SPDS been a useful tool in monitoring plant safety status under operating conditions? If not, why?

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. 4) Is there adequate documentation for using and interpreting the SPDS displays? Is this documentation readily available in the control room?

5) Have you been adequately trained in the use and interpretation of SPDS? If not, what improvements should be implemented?

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6) Which' control room staff members use the SPDS during your shift: Shift Engineer, Shift Foreman, Reactor Operator, Shift Control Room Engineer? Who would use it during an emergency?

Location

7) Is-the SPDS visible from the areas of the control room where you need to have the information displayed? If not, where should it be located?

8) Does the screen (s) block or distract from the view of other controls or displays? If so, which ones?

9) Are there enough CRT displays dedicated to the SPDS?

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10) Are there any lighting problems (i.e., glare, reflections)  ;

that impair use of the SPDS? Can the SPDS be used '

effectively under emergency lighting conditions?

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1 Interviewee No. j Parameter selection l 11) Does the'SPDS supply the appropriate information necessary to detect and monitor abnormal plant conditions?

If not, why?z 12)-Please identify.any presently-displayed parameters that-you do not think are useful. Also, identify any _

parameters that you think should be added.to.the display ~

(consider the requirements of Reg. Guide 1.97). In what format.should these proposed parameters ~be displayed?

Display characteristics

13) Is the format used to display.the plant parameters easy to. interpret? Can you visually detect changes in safety-status quickly and accurately? Please list any features that are currently well designed, as well as any.

suggestions for improvement.

14) Is color used effectively in the SPDS? Is the use of color in the SPDS consistent with the use of color elsewhere in the control room?

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15) Are validated and invalidated data easily distinguishable from one another on the SPDS display?

16) Are the scales and alarm limits shown on the iconics at i appropriate levels? Does the display alert the operator to parameters that may be trending toward an alarm condition before the alarm set point is reached?

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17) Are the SPDS alarm set points the same as those for corresponding annunciator system alarms? Does the SPDS effectively supplement the annunciator system?

18) Is the SPDS display labeled appropriately? Are abbreviations and terms consistent with those used elsewhere in the control room?

19) Do you have any additional comments or suggestions regarding the SPDS?

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J APPENDIX B PERSONNEL DEMOGRAPHICS OF BYRON AND BRAIDWOOD STATIONS SPDS INTERVIEW PARTICIPANTS B-1 a .-..- - --

Personnel Demographics of Bryon/Braidwood Stations SPDS Interview Participants 1.1 Current Job Classifications Twelve control room personnel at Byron and Braidwood Stations were interviewed during April 1987 Table 1 lists the current CECO positions and the number of interviewees holding those positions.

Table 1. Job Classification of Interviewees Shift Engineer 3 Shift Foreman 3 SCRE 4 Engineering Assistant i Operating Staff 1

1. 2 Nuclear Experience The distribution of experience of interviewees with CECO, in the nuclear industry, and in their present positions is presented in Table 2.

Table 2. Experience of Interviewees LENGTH OF EMPLOYMENT I O-5yr 6-10yr 11-15yr 16-2Oyr N

N T U E With Ceco 4 7 1 O M R B V In Nuclear E I Industry 2 6 4 0 R E W

O E LENGTH OF EMPLOYMENT F E O-6mo 7-12mo 1-3yr 4-7yr S

In Present Position 2 0 9 1 B-2

The average tenure of interviewees is 1.57 years in their current positions, 7.04 years at CECO, and 9.38 years in the nuclear . industry..

Of the twelve personnel interviewed, one indicated that he held an Reactor Operator's (RO) license.

Of the twelve personnel interviewed, eleven (92%) indicated that they held'the Senior Reactor Operator's (SRO) license. .  ;

The average time for holding the,SRO license was 0.83 years.  !

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9 APPENDIX C FINDINGS IDENTIFIED BY THE BYRON AND BRAIDWOOD STATIONS SPDS PERSONNEL INTERVIEW l

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Areas of Concern Identified by the Byron /Braidwood Stations SPDS Personnel Interview Area 1. GUIDELINE NO.: 5.3.1.2.A/NUREG-0800 l FINDING: The SPDS display does not function properly at the simulator. Therefore, operator training at the simulator does not include practice with transient events using the SPDS.

DISPOSITION: Upgrade on the Process computer is n currently being performed at the Byron /Braidwood simulator. These improvements will enhance the operation of the SPDS.

IMPLEMENTATION: 12-31-88 C-2

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l Areas of Concern Identified i by the Byron /Braidwood Stations SPDS Personnel Interview Area 2. GUIDELINE NO.: 5.7.2/NUREG-0800

, 5.3.1.2.A/NUREG-0800 0:NDING: The respondents were unsure as to whether the SPDS alarm setpoints corresponded to  !

the annunciator system alarms.

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[ISPOSITION: Operational limits of the plant i i'ere used to determine SPDS setpoints independent of annunciator setpoints. SPDS setpoints are covered in. training, v:

IMPLEMENTATION: Accept as is O-3 .

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Areas of Concern Identified by the Byron /Braidwood Stations SPDS Personnel Interview

. Area 3. GUIDELINE'NO.: 5.1.3.3.1.E/NUREG-0800 ~

5.1.3.3.1.B/NUREG-0800 5.1.3.3.1.D/NUREG-0800 5.3.1.2.A/NUREG-0800 FINDING: Most of the respondents were unaware of- )

the means by which validated and invalidated data f could be distinguished on the SPDS display.

DISPOSITION: An updated SPDS. Training Guide will be provided to the Training Department. The above concern will be addressed in the SPDS guide.-

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IMPLEMENTATION: 12-31-88 O-4

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Areas of-Concern Identified by the Byron /Braidwood Stations SPDS Personnel Interview r

Area 4. GUIDELINE NO. 5.2.1.2.A/NUREG-0800 FINDING: The' location of an SPDS'in the shift supervisor's office ~was. suggested.

DISPOSITION: The SPDS is properly displayed in the control room and the TSC, where they are needed.

I IMPLEMENTATION: Accept as is l

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Areas of Concern Identified by'the Byron /Braidwood Stations SPDS Personnel Interview Area 5.- GUIDELINE NO.: 5.3.1.2.A/NUREG-0800.

FINDING: A.magority of the respondents felt that the training they had received in the use of the SPDS was not adequate. For example, respondents indicated there was little formal training and that  !

the simulator's SPDS was often not functional. )

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DISPOSITION: Operators are trained on the various computcr systems and their experience.on these systems has increased.resulting in improved performance.

An updated SPDS Training. Guide will be provided to the

. Training Department.

Also, the simulator's SPDS is i currently being enhanced.

t. 3 IMPLEMENTATION: 12-31-88 i

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I Areas of Concern Identified by the Byron /Braidwood Stations SPDS Personnel Interview l

Area 6. GUIDELINE NO.: S.3.1.2.B/NUREG-0800 7.'. 8.A.1/NUREG-0700 7.1.8.A.3/NUREG-0700 7.1.8.A.4.C/NUREG-0700 FINDINGS. A number of respondents we.2 not aware of the existence of SPDS documentation (e.g.

procedures).

DISPOSITIONS. Operators are trained on the various computer systems and their experience on these systems has. increased resulting in improved performance.

An updated SPDS Training Guide will be provided to the Training Department. Also, the simulator's SPDS is currently being enhanced.

IMPLEMENTATION: 12-31-88 C-7

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APPENDIX D FINDINGS IDENTIFIED BY THE BYRON AND BRAIDWOOD STATIONS SPDS CHECKLIST REVIEW l

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l Areas of Concern Identified by the Byron /Braidwood Stations SPDS Checklist ".eview Area 7. GUIDELINE NO.: 5.2.1.2.C/HUREG-0800 7.2.1.8/NUREG-0700 FINDING: There is glare on the CRT screens used for the SPDS displays on PMO5J.

e DISPOSITION: Anti-glare screens.will be installed on the CRTs mounted on the PMO5J panel.

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IMPLEMENTATION: By completion of the 1st refueling outage D-2

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L-Areas of Concern Identified by the' Byron /Braidwood Stations SPDS Checklist Review Area 8. GUIDELINE NO.: 5.3.1.2.B/NUREG-0800 7.1.8.A.1/NUREG-0700 7.1.8.A.3/NUREG-0700

)

7.1.8.A.4.C/NUREG-0700 7.1.8.A.5.A/NUREG-0700 FINDING: A SPDS User's guide is not available in i

the control room. Such a reference would be useful to aid operators in interpreting the SPDS displays.

DISPOSITION: An SPDS User's Guide is not needed in the control room. The iconic display's shape alerts operators to possible plant deviations from normal. Procedures and training' direct operators to review appropriate control board instrumentation. In addition, the FSAR (App. E-17) lists and discusses SPDS setpoints.

IMPLEMENTATION: Accept as is D-3

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L Areas of Concern Identified l- by the Byron /Braidwood Stations SPDS Checklist Review Area 9. GyIDELINE NO.: 5.4.2.1.A/NUREG-0800 7.2.7.L.3/NUREG-0700 5.1.6.C.2/NUREG-0700 FINDING: Color coding is not used on the SPDS displays to indicate the approach to unsafe operation. For example, a color is not designated to represent " Caution" or " Approaching Critical Value".

i DISPOSITION: The deviation from reference of the iconic provides a cue that values are approaching abnormal conditions. Also, reference values and actual values are indicat-ed on the SPDS display.

IMPLEMENTATION: Accept as is D-4

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Areas of Concern Identified L by the Byron /Braidwood Stations SPDS Checklist Review Area 10. GUIDELINE NO.: 5.7.1.A/NUREG-0800 5.7.1.B/NUREG-0800 S.7.2/NUREG-0800 FINDING: The SPDS does not include a distinct audible sound to indicate an abnormal operating condition. Operators rely on annunciator alarms for auditory cues. l DISPOSITION: All SPDS parameters are included in the annunciator system. Additional auditory alarms may provide unneeded redundancy.

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l IMPLEMENTATION: Accept as is l

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Areas of Concern Identified by the Byron /Braidwood Stations SPDS Checklist Review Area 11. GUIDELINE NO.: 7.1.5.D.2/NUREG-0700 FINDING: The SPDS mode selection key is not readily distinguishable from other keys on the computer console. The key is marked " Misc Select" and is used to call up a menu which includes the SPDS displays.

DISPOSITION: The procedure required to select the appropriate range iconic will be addressed in the updated SPDS Training Guide. SPDS iconic selection is done by a series of menu-driven panels.

IMPLEMENTATION: 12-31-88 D-6 l

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!. Areas of Concern Identified by the Byron /Braidwood-Stations SPDS Checklist' Review Area 12. ' GUIDELINE'NO. . 5.1.4.1.A/NUREG-0800 FINDING: The SPDS display does not indicate-trends of each TSPDS variable. For example, there

.are no arrows or color codes on the iconics to

' indicate the, direction of changes.

DISPOSITION . . Secondary displays are available through the Process Computer which allows 30 minute trending for.SPDS parameters.

IMPLEMENTATION: Accept as im D-7

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Areas of Concern Identified by the Byron /Braidwood Stations SPDS Check' list Review Area'13. GUIDELINE NO.: 5.6.A.1/NUREG-0800 FINDING: The SPDS does not provide function keys for the recall of data. 'For example, the SPDS does not allow operators to directly retrieve specific data.

DISPOSITION: Operators are able to access all relevant data through the Prime and Process computers. (

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IMPLEMENTATION: Accept as is D-8 I

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Areas of Concern Identified by the Byron /Braidwood Stations SPDS Checklist Review Area 14. GUIDELINE NO.: 6.2.3.1.B/NUREG-0800 7.2.7.K.2/NUREG-0700 5.1.6.C.1/NUREG-0700 5.1.6.C.2/NUREG-0700 5.1.6.D.1/NUREG-0700 5.1.6.D.3/NUREG-0700 6.2.3.1.C.1/NUREG-0800 6.2.3.1.D.2/NUREG-0800 FINDING: Colors on the SPDS do not consistently conform to a color code. For example, amber and cyan may both be used to indicate normal conditions.

DISPOSITION: The SPDS does conform to a color standard used by Computer Systems. That standard is in part predicated on distinguish-able colors able to be generated on the CRTs.

The Judicial use of amber for the iconic outline was previously accepted by the NRC in a July 1985 Audit Report.

IMPLEMENTATION: Accept as is  ;

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L Areas of Concern Identified j by-the Byron /Braidwood Stations SPDS Checklist Review l

Arer 15. GUIDELINE NO.: 7.2.2.6.2/NUREG-0700 FINDING: .The SPDS displays use a 5 x 7 dot matrix for characters. The guideline suggests a 7 x 9 dot matrix.

DISPOSITION: Characters are readable. The 5 x 7 dot matrix is used for engineering units.

All active analog outputs are displayed in 10 x 14 dot matrixs which exceed the guideline.

I IMPLEMENTATION: Accept as is D-10

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, Areas of Concern Identified by the Byron /Braidwood Stations SPDS Checklist Review Area 16. GUIDELINE NO.: '6.2.3.1.D.1/NUREG-0800 5.1.6,0.2/NUREG-0700 7.2.7.K.1/NUREG-0700 6.2.3.1.B/NUREG-0800 FINDING: The colors used on the SPDS are not consistent in meaning with other color codes in the-control room. For example, amber is used to denote caution in the control room, but is used to indicate the active vector (which can be neutral) in the SPDS.

DISPOSITION: The resolution to this item is being evaluated as per the June 23, 1987 telephone conference with the NRC. It has been agreed upon that this resolution will be transmitted to the NRC by August 31, 1987.

IMPLEMENTATION:

D-11

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4 Areas of Concern Identified by the Byron /Braidwood Stations SPDS Checklist Review Area 17. GUIDELINE NO.: 6.2.3.1.C/NUREG-0800 FINDING: The first color change in the SPDS color code does not alert the operator that a variable is outside its normal range and does not yet represent a serious problem. The first color change is from.

cfan to magenta or to red; magenta and red indicate low and high values, respectively. Currently, operators rely only on the shape of the iconic to determine that a variable value is out of normal range.

DISPOSITION: Deviations in active parameters from the reference parameters are represented as points plotted on the normalized scales drawn between fixed reference and high and low limit points. Color would do little to enhance understanding and may interfere with recognition of the SPDS display.

IMPLEMENTATION: Accept as is D-12

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Areas of Concern Identified by the Byron /Braidwood Stations SPDS Checklist Review l

Area 18.- GUIDELINE NO.: 5.1.3.3.1.D/NUREG-0800 l FINDING: Operating procedures do not provide I j

guidance for treatment of invalid data for the l SPDS. Such a procedure would serve to sid .

operators in the correct interpretation of the SPDS l display. l DISPOSITION: An updated SPDS Training Guide will be provided to the Training Department. The treatment of invalidated data will be addressed in the SPDS Training Guide.

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l IMPLEMENTATION: 12-31-88 D-13

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APPENDIX ~E AREAS OF CONCERN IDENTIFIED

~BY'THE BYRON /BRAIDWOOD STATIONS SPDS/DCRDR TASK ANALYSIS REVIEW s.

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e Areas of Concern Identified by the Byron /Braidwood Stations SPDS/DCRDR Task Analysis Review Area 19. GUIDELINE NO.: 5.1.1.A.1/NUREG-0700 FINDING: During the review of the DCRDR Task l Analysis Data Base, Surge Tank Level was identified as a parameter that should be displayed on the SPDS. This is the water supply for the Component Cooling Syatem whose status can be determined from the tank level. If component Cooling is lost there would be no cooling to unit pumps, and the RHR Heat Exchanger, the RCP's, and RCP Thermal Barrier Flow would be lost.

DISPOSITION: A modification is currently being designed to provide for automatic make-up to the CC Surge Tank when a low level setpoint is reached. Upon implementation of this modification, Surge Tank-level will only fluctuate minutely between two narrowly defined setpoints. The l

presentation of level to the operator displayed on_the SPDS would not provide any significant information since the system will be automatically  ;

maintained.

IMPLEMENTATION: By completion of the 1st refueling outage E-2

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Areas of Concern Identified by the Byron /Braidwood Stations SPDS/DCRDR Task Analysis Review

' Area 20. GUIDELINE NO.: 5.1.1.A.1/NUREG-0700 FINDING: During the review of the DCRDR Task Analysis Data Base the following parameters were identified as desirable.to have available on a secondary SPDS display. Ready access to this information would enhance normal. operations and may facilitate operations during a transient event.

1) Exhaust Hood Pressure

2) FW Pump Suction Pressure

3) VCT Level

4) IA Header Pressure

5) DG Output Current, Watts, Vars & Frequency

6) ESF Bus Voltage

7) HD Pump Amps

8) CD/CB Pump Amps

9) CW Pump Amps

10) CD and CB Pump Discharge Pressures

11) FW and MS Flow

12) Hot and Cold Leg Temperatures DISPOSITION: The use of secondary or backup displays to the primary SPDS display is not addressed as a requirement in NUREG-0737, Supplement i for SPDS. This information is adequately displayed on other control room displays.

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IMPLEMENTATION: Accept as is E-3

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Areas Of Concern Identified ,

t by the Byron /Braidwood Stations SPDS/DCRDR Task Analysis Review Area 21. GUIDELIslE NO.: '5.1.1.A.1/NUREG-0700 FINDING: During the review of the DCRDR Task Analysis Data Base, Steam Generator Steam Pressure was identified as a parameter that should be displayed on the SPDS. This pressure is inversely proportional to power and tells the operator the status of the integrity of the Secondary Heat Removal side of the plant.

DISPOSITION: Steam Generator Steam Pressure is not a primary reactor safety parameter. The use of secondary or backup displays is not addressed as a i requireinent in NUREG-0737, Supplement i for SPDS.

This information is adequately displayed on other control room displays.

IMPLEMENTATION: Accept as is E-4 l

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