ML19325C145
| ML19325C145 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/10/1989 |
| From: | SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| Shared Package | |
| ML19325C146 | List: |
| References | |
| CON-FIN-D-1131 SAIC-89-1130, TAC-M56107, TAC-M56108, NUDOCS 8907190108 | |
| Download: ML19325C145 (45) | |
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t ATTACHMENT 1 (to Staff Evaluation) i i
SAIC 89/1130 f TECNNICAL EVALUATION REPORT FOR ;
THE DETAILED CONTROL ROOM DESIGN REVIEW i AT CAROLINA POWER AND LIGHT COMPANY'S '
., BRUNSWICK STEAM ELECTRIC PLANT, UNITS I and 2 !
TAC NOS. M56107, M56108 5 i
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l Prepared for:
U.S. Nuclear Regulatory Comission g
Washington, D.C. 20555 Contract NRC-03-87-029 d
Task Order No. 35 4F
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TABLE OF CONTENTS Section g 1.0 INTR 000CTION.............................................., 1 1.1 Background............................................ I 1.2 Audit Agenda and Participants......................... 2 2.0 EVALUATION................................................. 2 2.1 Establishment of a Qualified Multidisci Review Team...................................plinary ............... 3 2.2 System Function and Task Analysis..................... 3 2.3 Comparison of Display and Control Requirements With a Control Roor Inventory................................ 5 2.4 Cont rol Room Survey . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2.5 Assessment of Human Engineering Discrepancies (HEDs)
To Determine Which Are Significant and Should be Corrected............................................. 6 2.6 Selection of Design Improvements...................... p 2.7 Verification that Selected Design Improvements Will Provide the Necessary Correction...................... 9 2.8 Verification the Improvements Will Not Introduce New HED3.................................................. 11 2.9 Coordination of Control Room Improvements With Changes From Other Programs, such as the Safety Parameter Display System, Operator Training, Regulatory Guide 1.97 Instrumentation, and U Procedures. . . . . . . . . . ................................
. .pgraded Emergency Operating 12
3.0 CONCLUSION
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4.0 REFERENCES
................................................. 14 Attachment 1 Audit Agenda Attachment 2 List of Audit Participants Attachment 3 Potential HEDs Identified in the Audit Team's Walkthrough of E0P-04-RRCP i
TECHNICAL EVALUA710N REPORT FOR t THE DETAILED CONTROL ROOM DESIGN REVIEW -
AT CAROLINA POWER AND LIGHT COMPANY'S !
BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 I i
1.0 INTRODUCTION
This report documents the findings of a pre-implementation audit of the Carolina Power and Light Company's Detailed Control Room Design Review.
(DCRDR) at the Brunswick Steam Electric Plant. The audit was conducted by the Nuclear Regulatory Comission (NRC) during a site visit May 15 through May 18, 1989.
The purposes of the audit were:
o To assess the licensee's progress toward completing the nine DCRDR requirements stated in NUREG-0737, Supplement 1 (Reference 1). -
o To discuss the licensee's plans and projected schedules for '
completing the DCRDR program at the Brunswick Steam Electric Plant.
The audit agenda is provided as Attachment I to this report and a list of audit meeting participants is provided in Attachment 2.
1.1 Background
The following is a chronological list of milestones in the Brunswick L Steam Electric Plant DCRDR:
1981 The licensee conducted a review of the Brunswick Unit I and 2 control rooms using the criteria in NUREG/CR-1580.
i 12/84 Program plan for conducting the DCRDR submitted to NRC by licensee (Reference 2).
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k 5/85 NRC staff comments on the program plan provided to licensee '
(Reference 3), i 9/85 In Progress audit of the Bruaswick DCRDR, conducted by NRC.
12/85 Updated program plan for conducting the DCRDR submitted to NRC by the licensee (Reference 4).
12/86 Final DCRDR Sumary Report submitted to NRC (Reference 5).
6/87 Revision 1 of the DCRDR Final Sumary Report submitted, including an updated implementation schedule, (Reference 6).
5/89 NRC conducted the Pre-implementation audit of the Brunswick '
! DCRDR.
t 1.2 Audit Agenda and Participants l' .
The licensee provided an opening sumary of DCRDR program status, work in progress and work to be done. Thereafter, each of the nine DCRDR requirements of NUREG 0737, Supplement 1, was reviewed using the guidance ,
provided in Section 18.1 of NURD-0800 The Standard Review Plan (Reference 7); and in NUREG 0700, Guidelines For Control' Room Design Reviews (Reference !
8). A technical discussion of findings was conducted with the licensee's DCRDR project team. In addition, the findings were sumarized in a formal exit briefing given by the NRC audit team leader.
The audit team consisted of a NRC team leader and NRC contractors from l Science Applications International Corporation (SAIC) and Comex Corporation, I representing the disciplines of human factors engineering and nuclear operations. The licensee's team included members from several divisions within Carolina Power and Light Company.
l 2.0 EVALUATION
! In the following sections the status of the Brunswick Steam Electric Plant DCRDR is evaluated with respect to each of the nine DCRDR requirements stated in NUREG-0737, Supplement 1.
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l 2.1 Establishment of a Qualified Multidisciplinary Review Team l
The organization for conduct of a successful DCRDR can vary widely but l 1s expected to conform to some general criteria. Overall administrative '
leadership, should be provided by a utility employee, who should be given l
sufficient authority to ensure that the DCRDR team is able to carry out its mission. A -core group of specialists in the fields of human factors engineering and nuclear operations and engineering are expected to participate with assistance as required from personnel in other disciplines.
Human factors expertise should be included in the staffing of the technical '
tasks. Finally, the DCRDR team should receive an orientation briefing on the DCRDR purpose and objectives which contribute to the success of the DCRDR. NUREG 0800, Section 18.1, Appendix A describes criteria for the multidisciplinary review team in more detail.
The DCRDR team was managed by a licensee representative. The DCRDR team consisted of individuals with expertise in the areas of instrumentation and control engineering, nuclear systems engineering, nuclear power plant" operations, training, licensing, and human factors engineering. Human factors engineering contractor support was provided by RMS Associates and Essex Corporation. The team was largely still intact at the time of the audit, with the DCRDR in the corrective action phase. The team is still involved in reassessments of Human Engineering Discrepancies (HEDs), and the finalizing of corrective action plans now in progress. .
It was the audit team's judgment that the licensee met the NUREG-0737, Supplement I requirement for establishment of a qualified multidisciplinary audit team.
2.2 System Function and Task Analysis The purpose of the system function and task analysis is to identify the control room operator's tasks during emergency operations and to determine the information and control capabilities the operators need in the control room to perform those tasks. An acceptable process for conducting the task analysis is as follows:
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- 1. Analyze the functions performed by plant systems in responding to transients' and accidents in order to identify and describe those ;
tasks operators are expected to perform.
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7 2. For each task identified in Item 1 above, determine the information'(e.g., parameter, value, status) which signals the need to perform the task, the control capabilities needed to '
perform the task, and the feedback information needed to monitor I task performance. '
- 3. Analyze the information and control capability needs identified in Item 2 above to determine appropriate characteristics for displays and controls to satisfy those needs.
The licensee conducted the DCRDR system function and task analysis in-coordination with the emergency operatir.g procedures (EOP) upgrade program.
The task analysis was based on the Boiling Water Reactors owners Group; generic Emergency Procedures Guidelines (EPGs), !!evision 3, the Brunswick Plant-Specific Technical Guidelines developed from the generic EPGs, and the Brunswick symptom based E0Ps. The generic Graphic Display System requirements document, developed by EPRI with Owners Group participation, was also used in the task analysis.
It was the audit team's judgment that the licensee has met the NUREG-0737, Supplement i requirement for a system function and task analysis.
However, the licensee has since upgraded their E0Ps to include Revision 4 of the generic EPGs. As a result of upgrading their E0Ps to Revision 4 of the EPGs, several changes have occurred to operator information and control requirements and these changes have not been analyzed to determine their effect on the control room instrumentation. The audit team recommended evaluating the changes resulting from upgrading E0Ps to Revision 4 of the EPG's and incorporating any corresponding changes to the control room instrumentation.
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.,- q 2.3 Comparison of Display and Control Requirements with a Control Room Inventory '
The purpose of comparing display and control requirements to a control ;
room inventory is to determine the availability and suitability of displays ,
and controls required to perform the emergency operating procedures. The success of this element depends on the quality of the system function and ,
task analysis and the control room inventory. The control room inventory should be a complete representation of displays and controls currently in :
the control room. The inventory should include appropriate characteristics '
of current displays and controls to allow meaningful comparison to the ,
function and task analysis. Unavailable or unsuitable displays and controls !
should be documented as human engineering discrepancies (HEDs).
The licenses- documented the operator information and control requirements along with instrumentation and control characteristics requirements en task analysis fonns. These forms were then used to verify i availability and suitability of controls and displays by comparison to both' the inventory .and to actual control room equipment during E0P walkdowns.
' This ~ included consideration of whether the equipment exhibited the proper characteristics as well as whether it met appropriate human engineering guidelines. The results of the licensee's evaluation were documented in Appendix A-14 to the Final Sunnary Report.
In order to test the licensee's results, the audit team conducted a
) control room walkdown of the Level / Power Control procedure (EOP-01-LPC) and the Radioactivity Release Control procedure (EOP-04 RRCP) to identify potential HEDs that should have been identified by the Brunswick DCRDR. In the walkdown of E0P-01-LPC, the audit team identified no HEDs that had not been identified and addressed by the licensee. The audit team identified six potential HEDs during the walkdown of E0P-04-RRCP (see Attachment 3).
The licensee was able to demonstrate that they had identified one of these g HEDs. The potential HEDs identified by the audit team involved i
discrepancies between the E0Ps and labeling in the control rooms.
The Radioactivity Release E0P is one procedure that has been updated to Revision 4 of the generic EPGs. The discrepancies noted in the walkthrough of this procedure underscore the need to update the task analysis and l 5 '
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perform a comparison to control room comoonents for Revision 4 and future revisions.
' It was the audit team's judgment that the licensee met the NUREG-0737, Supplement 1 requirement for a comparison of display and control requirements with a control room inventory.
2.4 Control Room Survey The key to a successful control room survey is a systematic comparison of the control room to accepted human engineering guidelines and human factors principles. One' accepted set of human engineering guidelines is provided in NUREG-0700; however, other accepted humar, factors standards may be chosen. Discrepancies should be documented as HEDs.
Tife objective of the licensee's control room survey was to identify any u characteristics of instruments, equipment, layout, and ambient conditions that did'not conform to good human engineering practice. Survey Task Plans were used which incorporated the human engineering criteria from HUREG-0700. ;
l It was the audit team's judgment that the licensee met the NUREG 0737, Supplement I requirement for a control room survey.
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2.5 Assessment of Human Engineering Discrepancies (HEDs) to Determine Which i Are Significant and Should be Corrected Based on the guidance of NUREG-0700 and the requirements of NUREG-0737, Supplement 1, all HEDs should be assessed for significance. The potential for operator error and the consequence of that error in terms of plant l
safety should be systenatically considered in the assessment. Both the
- j. individual and aggregate effects of HEDs should be considered. The result of the assessment process is a determination of which HEDs should be corracted because of their potential impact on plant safety. Decisions on whether HEDs are safety-significant should not be compromised by j- consideration of such issues as the means and potential costs of correcting HEDs.
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J A description of the licensee's assessment process was provided in l Section 4 of the Final Summary Report. The assessment was performed by a suitably qualified n,ultidisciplinary team. The process was designed to 1 prioritize HEDs based on estimations of the potential for error and f consequences of error. A numerical value was assigned to identify the !
priority of each HED. Corrective action recommendations were developed and documented by the assessment team. When it was decided that no corrective i action would be required, the justification was documented. Review of )
several HEDs sampled during the audit indicated that the licensee followed the process outlined in the Final Sumary Report.
Recently the licensee has reevaluated the need to correct several HEDs.
Among these are five Priority 2 HEDs included in the E0P Instrumentation l- Project: HEDs 206X-5092, 5093, 5094, 5096 and 5097. Tha justifications for l not correcting two of these HEDs (5096 and 5097) were judged by the audit team to be acceptable. The remaining three HEDs are still being re-evaluated by the licensee. A decision about correcting these HEDs is i expected to be mcde in June 1989, and the licensee will inform the NRC of any changes to the original corrective actions proposed for these HEDs.
The licensee reported that the need to correct two additional HEDs has been reevaluated. They are HEDs 206X-2106 and 2115 in the Annunciator Project.
- HED 2CSX 2115 concerns annunciator tiles which are not functionally grouped. This HED was originally assessed as Priority 3, defined as involving significant error potential but an insignificant consequence of error. The tile relocation task was cancelad based on the large number of procedures which would have been affected, and the extremely :omplex wiring changes which would be required. Although this task has been canceled, the licensee indicated that future additions of annunciators will be placed in the proper functional position.
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- l. HED 206X-2106 concerns tha global design of the annunciator alarm L response controls. This was originally assessed as a Priority 2 HED, l- defined as involving both significant probability of error and a significant L consequence of error. Under the original assessment recommendation, the l annenciator response " joy-:. ticks" were to be separated into zones so annun-l l 7 1'
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1 ciators could only be silenced, acknowledged, reset and tested from joy-sticks located in proximity to the annunciators effected. As built, there are about eight joy-sticks on the boards of each control room. Any of the eight joy sticks can be used to silence and acknowledge any annunciator in the control room.
The licensee decided to cancel the joy stick zoning task. This decision was reportedly based on cost and operator objections to the modification. The audit team noted an incongruity in the fact that operators may have identified the HED and then objected to the correction.
Discussions with onsite NRC personnel indicate that there has been at least one recent occurrence where operators silenced an annunciator without noting i it, thereby not realizing that a piece of Emergency Core Cooling System
.(ECCS) had been activated and then tripped. In the audit team's judgment, there is a significant probability of an operator error occurring because an I alarm can be silenced and acknowledged from a location where it cannot be read, combined with the fact that one joy stick manipulation can silence or acknowledge all alarms on all control boards. This was the DCRDR assessmenf team's original judgment as well. l It was stated that an administrattve !
control has been implemented requiring the operator to go to an annunciator and identify it before silencing / acknowledging. It is the audit team's judgment that administrative control may not be reliable in an emergency I situation.
l The addit team determined that the licensee has no't met the NUREG 0737, L
Supplement I requirement for assessment of HEDs to determine which are significant and should be corrected. To complete this requirement, the licensee should complete the ongoing reevaluation of corrective actions for the three E0P instrumentation HEDs identified above. The licensee should also reconsider its decision about HED 206X-2106 concerning annunciator control zoning. In both cases, the licensee should ensure that the decision criteria of error probability and error consequences are properly taken into L account, as required by the licensee's DCRDR assessment methodology, and that cost does not become the over-riding factor.
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P 2.6 Selection of Design Improvements '
The purpose of selecting design improvements is to determine, corrections to HEDs identified in the review phase of the DCRDR. " election of design improvements should include a systematic process for the development and comparison of alternative means of resolving HEDs.
Furthermore, according to NUREG-0737, Supplement 1, the licensee should document all of the proposed control room changes.
1 Approximately one third of the proposed corrective actions identified in the 1986 Final Summary Report and in Revision 1 to that report were ;
installed in the Brunswick control rooms at the time of the audit. The NRC i audit team reviewed several implemented changes. Enhancement modifications, i
including new labeling and annunciator priority coding, were judged in accordance with NUREG-0700 guidance. Design changes such as new recorders and meters and relocation of redundant reactor vessel instrumentation were also judged to be in accordance with NUREG-0700 guidance. No " Priority 1" l safety significant HEDs remain to be corrected in the control room.'
I Approximately half of the " Priority 2" safety significant HEDs have been
- . corrected. The licensee has scheduled the corpletion of all DCRDR related l modifications for both units in the 1989 1992 time frame. The audit team I reviewed selected work packages for these modifications. There are l- approximately 56 work packages in various stages of preparation.
l It was the audit team's judgment that the licensee has met the NUREG-0737, Supplement I requirement for selection of design improvements.
2.7 Verification that Selected Design Improvements Will Provide the Necessa.ry Correction l A key criterion of DCRDR success is a consistent, coherent, and l effective interface between the operator and the control room. This l criterion may be met by effectively executing the processes of selection of 1-design improvements, verifict. tion that selecteo improvements will provide the necessary correction, and verification that the improvements will not introduce new HEDs. According to NUREG-0800, techniquos for the l verification process might include resurveys of panels, applied experiments, l engineering analyses, environmental surveys, and operator interviews. The l-9
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consistency, coherence, and effectiveness of the entire operator-control room interface are important to operator performance.
The verification of corrective actions in the Brunswick DCRDR was '
identified as the responsibility of the HED assessment tesm. They addressed this requirement as part of the processes of selecting corrective actions and developing detailed corrective action plans in the various assessment follow up projects. The licensee identified the following steps taken to ensure that corrective actions would resolve the identified problems:
o Grouping. of HEDs that address the same type of problem to enture integration ar.d consistency of resolution.
o Grouping of HEDs that addressed the same compenent or type of component to ensure consistency of resolution.
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o Evaluation of proposed corrective actions against applicable NUREG-0700 criteria.
Several control room human factors engineering standards were L developed. They cover: labeling design, acronyms and abbreviations, color 3 L coding, and zone coding. These standards and other human factors guidelines
[ have been incorporated into a Human Factors Design Guide and a site specification, " Human Factors Engineering fcr Control Panel Modifications" (Specification No. 170-001). This guidance was reviewed durir.g the audit and found to be appropriate.
l In Jaruary 1989, a new, corporate wide Nuclear Engineering Department procedure was issued which governs all plant modifications, including control room modifications. This procedure requires consideration of NUREG-0737, Supplement 1, requirements as applicable in the preparation of design modifications. It states that the site Human Factors Engineering Guide should be used as applicable. It requires a human factors engineering review, when applicable, as part of the process of developing a modification. This is among the review responsibilities assigned to Operations. The modification procedure requires conducting a post-implementation walkdown, but does not specify attention to human factors 10
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engineering issues during the walkdcwn. This procedure does not require separate human factors engineering signoff en modifications; the overall i
Operations signoff is presumed to take human factors issues into consideration where applicable. The audit team reconnended making the human .
factors engineering requirements in this procedure more evident and more explicit, and requiring a separate human factors signoff when applicable to I a modification.
The licensee stated that the DCRDR team has taken, and will continue to take, an active role in ensuring that all DCRDR corrective actions are properly implemented to resolve the identified problems without creating new HEDs. Information was presented during the audit to support this statement, j A single coordinator has been assigned to manage all DCRDR related l modification packages. Another individual has been assigned responsibility to ensure that HED corrective action commitments are properly closed out.
The audit team reviewed documentation of implemented corrective actions '
which indicated that, to date, there has been appropriate post-i implementation follow-up to verify control room changes resulting from the l- DCRDR. In addition, audit team verified a sample of implemented corrective actions which indicated that the corrective action verification process has
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been conducted satisfactorily.
1 l-It was the audit team 's judgment that the licensee met the requirement of NUREG 0737, Supplement 1, for a verification that the selected design improvements will provide the necessary corre.: tion.
l 2.8 Verification that the Improvements Will Not Introduce New HEDs l
'T As discussed in Section 2.7 above, the licensee did have a process for verifying that the improvements will not introduce new HEDs when implemented. Therefore, it was the audit teem's judgment that the licensee has met the requirement of NUREG-0737, Supplement 1, for a verification that the improvements will not introduce new HEDs.
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2.9 Coordination of Control Room improvements with Changes From Other Programs, Such as the Safety Parameter Display System, Operator Training, Regulatory Guide 1.97 Instrumentation, and Upgraded Emergency l Operating Procedures Improvement of emergency response capability requires coordination of '
the DCRDR with other activities. Satisfying Regulatory Guide 1.97 i requirements and the addition of the Safety Parameter Display System (SPDS) :
necessitate modifications and additions to the control room. The modifications and additions should be specifically addressed by the DCRDR.
Exactly how the modifications are addressed depends on a number of factors ,
including the relative timing of the varicus emergency response capability upgrades. Regardless of the means of coordination, the result should be >
integration of Regulatory Guide 1.97 instrumentation and SPDS equipment into j a consistent, coherent, and effective control room interface with the i operators.
l l Management of the NURiG-0737 Supplement 1 initiatives at Brunswick was the responsibility of one individual. In addition, team members were .
assigned to work across projects. These management and staffing provisions g contributed to project coordination.
l The SPDS displays were developed by many ef the same people who worked on the DCRDR. The HED assessment team also supported development of the final SPDS displays. The human factors review of the SPDS was performed by (
L the lead human factors specialist for the DCRJR. Some DCRDR HEDs were resolved by incorporating data into both the SPDS and the E0Ps. For example, the Brunswick control rooms do not hava an integrated group i isolation status light display. To resolve this HED, a group isolation checklist was provided in the E0Ps and SPDS displays show isolation status at both a sumary level and in detail.
The site project coordinator for the DCRDR also had primary responsibility for the E0P upgrade. The DCRDR task analysis was based on the symptom-based E0Ps upgraded through Revision 3 of the Boiling Water Reactor Owners Group Emergency Procedure Guidelines. Seventeen HEDs were resolved by the E0P upgrade project. In addition, needs for additional instrumentation were identified in conjunction with the E0P upgrade project 12
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and became HEDs to be corrected in the DCRDR. The standard for acronyms and abbreviations that was developed in the DCRDR is being used in the procedures program and in the SPDS development effort.
Some HEDs were resolved by connunicating them to the training department. Training on panel modifications has been provided. The training simulator is being changed to maintain consistency with the control rooms.
Instrumentation upgrades, additions, or replacements to meet criteria in Regulatory Guide 1.97, Revision 2, and in NUREG-0737, Supplement 1, were made in the overall Emergency Response Capability project. Regulatory Guide 1.97 displays were reviewed in the DCRDR.
It was the audit team's judgment that the licensee met the requirement of NUREG 0737, Supplement 1, for coordination of the DCRDR with the development of the SPDS, upgraded E0Ps, operator training, and Regulatory Guicle 1.97.
3.0 CONCLUSION
The NRC conducted a Pre-implementation audit of Carolina Power and '
Light Company's Brunswick Steam Electric Plant Detailed Control Room Design Review during a site visit May 15 and May 18, 1989. The purposes of the audit were to assess the licensee's completion of the nine DCRDR requirements stated in NUREG-0737, Supplement I and to discuss the licensee's schedules for completing all corrective actions resultir.g from the program. It was the audit team's judgment that the licensee met eight of the nine DCRDR requirements.
The NUREG-0737, Supplement I requirement to assess all HEDs for safety significance and determine whether corrective action is needed was judged by the audit team to be incomplete. As discussed in Section 2.5 of this report, the licensee recently undertook reevaluation of the need to correct several HEDs. To complete the assessment requirement satisfactorily, the licensee should complete these reevaluations in a manner consistent with the assessment methodology defined in the licensee's Final Summary Report on the Brunswick DCRDR.
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4.0 REFERENCES
- 1. NUREG 0737, Supplement 1, " Clarification of TM1 Action Plan ,
Requirements," U.S. Nuclear Regulatory Comission, December 1982.
- 2. Control Room Design Review Detailed Program Plan and Implementation l
Guidelines for Carolina Power and Light Company's Brunswick Steam ;
Electric Plant Units 1 and 2. Carolina Power and Light Company, '
December 1984.
- 3. Coments on the Detailed Control Room Design Review Program Plan for ,
Carolina Power and Light Company's Brunswick Steam Electric Plant, Units 1 and 2. USNRC, May 1985. [
- 4. Update of Centrol Room Design Review Detailed Program Plan and Implementation Guidelines for Carolina Power and Light Company's
! Brunswick Steam Flectric Plant, Units 1 and 2. Carolina Power and,
. Light Company, December 1995.
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- 5. Brunswick Steam Electric Plant Control Room Design Review Final Sumary Report, Carolina Power and Light Company, December 1986. .
- 6. Brunswick Steam Electric Plant Control Room Design Review Final Sumary Report, Revision 1. Carolina Power and Light Company, June 1987.
- 7. NUREG 0800, " Standard Review Plan," Section 18.1, "Centrol Room," and I Appendix A. " Evaluation Criteria for Detailed Control Room Design L Review (DCRDR)," USNRC, September 1984.
- 8. NUREG-0700, " Guidelines for Control Room Design Reviews," USNRC, September 1981.
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ATTACHMENT 1 ( }o TER) !
AUDIT AGENDA 4 e
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ATTACHMENT 1 AUDIT AGENDA ()o 7"E d. )
M - May 15. 1989 1:30 pm NRC Entrance Briefing and Detailed Control Room Design Review (DCRDR) briefing 2:00 pm Licensee briefing on DCRDR program at Brunswick 2:45 pm Audit Controlteam control room walkdown or Radioactivity of the Reactor Pressure Vessel Release Control procedure (Access to control room needed; 2 licensed operators needed).
4:30 pm Audit team documentation of sample human engineering discrepancies identified during control room walkdown 5:00 pm End Day 1 Day 2 - Tuesday. May 16. 1989 '
8:00 am Presentation to licensee of findings from E0P walkthrough identified by audit team 8:30 am Review -of implemented and proposed DCRDR' related control room modificatioris including:
o Annunciator project modifications o E0P instrumentation project o Indicator upgrade project o Component relocation project 11:00 am Review of schedules for implementating any remaining safety significant HEDs.
11:30 am Review of coordination of DCRDR modifications with changes made in 1 other programs including:
o Safety Parameter Display System L o Regulatory Guide 1.97 instrumentation l o Upgraded E0Ps o Operator training 12:00 Lunch 1:00 pm Licensee discussion of how audit team's sample findings identified during E0P walkdown, were identified in their DCRDR. I 1
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2:00 pm NRC Caucus 3:00 pm NRC and licensee technical issue discussion and resolution l 5:00 pm End day 2 Day 3 - Wednesday. May 17. 1989 8:30 am NRC Safety Parameter Display System (SPDS) entrance briefing o SPDSs Generic Letter Status -
o Previous NRC findings regarding Brunswick SPDS 9:00 am Licensee briefing on SPDS program results to date 10:00 am SPDS Evaluation (Access to control room and TSC needed)
- 1. Parameter selection o Reactivity control o Core cooling and heat removal from the primary system o Reactor coolant system integrity ,
o Radioactivity control o Containment conditions
- 2. Continuous display of top level safety function information
- 3. Concise display of safety function information
- 4. Located convenient to control r'oom operator
- 5. High degree of reliability
- 6. Design incorporating human factors engineering
- 7. Procedures and training for SPDS operation
- 8. Electrical isolation L 12:30 Lunch 1:30 pm Operator interviews (S/S;STA;TRG instructor) 1- 3:00 pm NRC Caucus l
l 4:00 pm NRC/ Licensee SPDS technical issues discussion and resolution L 5:00 pm End day 3 l
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8:00 am - NRC/
necessary) Licensee discussion and resolution of DCPDR or SPOS issues (as -
. 10:30 am NRC/ Licensee management exit f
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ATTACHMENT 2 ie LISTOFAUDITPARTICIP(ANTS TER) q a e
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LIST OF AUDIT PARTICIPANTS ENTRANCE MEETING Attendees 5/15/89 i
H&tlE ORGANIZATION Sam Strickland OPS James Bongarra, Jr. NRC/NRR/DLPQ Gary Bethke NRC - Comex Barbara Paramore William H. Ruland NRC - SAIC (human factors)
NRC - SRI C.F. Blackmon, Jr. CP&L MGR - OPS William B. Geise Project Special, - Simulator Support George Barnes CP&L Operations :
Mike Williams BTU
- Mike Sawtschenko CP&L Operations Mike Beck OPS Ralph Sanders CP&L NED. Raleigh Randy Weiss CP&L NED, Raleigh Michael J. ' Pastva, Jr. CP&L Regulatory Compliance ,
Arnold W. Schmich CP&L NFS, Raleigh Wilbert May CP&L NED, Raleigh !
T.H. Wyllie CP&L - BNP L
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1 EXIT MEETING Attendees 5/18/89
. H&HE ORGANIZATION G.W. Bethke NRC - Comex Barbara Paramore NRC SAIC Mike Beck OPS i
Mike Williams TRANG I' Mike Sawtschenko OPS
- George Barnes OPS '
l Gene Eagle CP&L - TS Ralph Sanders CP&L, Raleigh K.E. Enzor p
. W. Levis NRC i
Joe Holder CP&L i
1
' T.H. Wyllie CP&L W.B. Geise CP&L J. O'Sullivan CP&L Walt Simpson CP&L Steve Callis CP&L ,
David Dcrsett CP&L David Rudoff QA/CP&L Arnold Schmich CP&L/NFS Michael Pastva CP&L Albort, May ' CP&L C.F. Blackmon, Jr. CP&L ,
R.E. Helme CP&L James Bongarra NRC/NRR L
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ATTACHMENT 3 ble NN) '
POTENTIAL HEDs IDENTIFIED IN THE AUDIT TEAM's WALKTHRODGH OF E0P 04 RRCP l
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I ATTACHMENT 3ha TER)
POTENTIAL HEDs IDENTIFIED IN THE AUDIT TEAM'S I WALKTHROUGH OF E0P 04 RRCP o The Brunswick Chemistry Department indicated that both the High i' and the High High alarm setpoints on the Process Offgas Vent Pipe (Plant Stack) were set the same at 4.2 E+5 uct/sec. The E0P entry )
condition uses a value of 2.94 E+5 uct/sec as ths High value. J This discrepancy between the E0P and the control room 'nstrument !
should be resolved. f o The E0P refers the operator to recorder D32 RR 4600 to read '
Process Off back panel, andgasprovides Vent Pipe This recorder is on a an radiation LED read level.
out in the proper engineering l units. The more accessible recorder on the front panel is labeled !
D12 R600, and reads in the base 10 logarithm of the engineering i
value(uct/sec). !
o The entry condition for Reactor Building Roof Vent Rad High (noble ;
gas release) has an E0P setpoint of 3400 CPM for unit 1, and 4200 '
CPM for unit 2. A paper label is attached to the actual unit 2 instrument indicating that the setpoint is 100,014 CPM. The unit, I instrument had no setpoint label, and appeared to be set close l to 4200 (based on potentiometer setting). The E0Ps, and the con- i trol room instruments should use the same setpoint and shoJ1d be "
labeled consistently, t
t-o An E0P entry condition for Service Water Effluent Rad High has !
setpoints in the E0P listed in units of CPM. The actual recorder !
i in the control room of CPS, but doesn't have the nitsu(012 RRthe604) is calibrated Control in units listed on instrument. room !
instrument 012-K605 in the control room is labeled in CPS. The i recorder should be labeled with units, and the E0Ps corrected to ;
reflect CPS versus CPM. The E0P problem had been previously iden- i i tified by CP&L. t l c Step RR/RB 9 of the E0P may require the operator to read *MSIV PIT Temperature'. The value is available on the back panels on the +
ECCS Leak Detection panel (1 821B 51). Obtaining the value :
L involves rotating a 19 position selector switch. The switch and instrument are not labeled 'MSIV PIT Temperature". Either the instrument should be labeled like the E0P, or the E0P corrected to i read like the instrument label. !
o The Process Reactor Building Vent Rad and Main Steam Line Rad I monitors on the back panels are not labeled with the instrument j designations (012 RM k609A & B, and 012 RM K603#. & B),
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I ATTACHMENT 2 (to Staff Evaluation) l I
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I HUMAN ENGINEERING DISCREPANCY l ANNUNCIATOR PROJECT (SAM STRICKLAND) ;
E0P INSTRUMENTATION PROJECT (ARNOLD SCHMICH) {
COMPONENT REM 0 vat PROJECT ,
INDICATOR UPGRADE PROJECT i COMP 0NENT RELOCATION PROJECT '
CONTROL ROOM CUNVENil0N PROJECT !
CONTROL ROOM HVAC PROJECT OFF-Gas FLON INSTRUMENT PROJECT !
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(4 llo8ED/14h )
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. i 6.3.3 THE COMPONENT REMOYAL PROJECT CONTAINs 18 HEDs. WHICH !
ARE tbVERED IN N PLANT MODIFICATIONS. INESE HEDs REMOVE UNNECESSARY COMPONENTS ON THE CONTROL BOARDS. !
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6.3.4 THE INDICATOR UPGRADE PROJECT CONTAINS 16 HEDS. WHICH l ARE COVERED IN SEVEN PLANT MODIFICATIONS. THESE HEDS ,
, ADDRESSED PHYSICAL CHARACTERISTICS OF INDICATORS SUCH !
AS, NUMBER SCALE PROGRESS 10NS j AND READABILITY OF !
INTERNAL SCALE LABELING. THE FUNCTIONAL ASPECTS OF METERS. AND LEGEND LIGHTS AND THEIR LABELS.
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6.3.5 THE COMPONENT RELOCATION PROJECT CONTAINS 3JHEDS, WHICH ARE COVERED IN 31 PLANT MODIFICATIONS. THESE PLANT MODIFICATIONS WILL FUNCTIONALLY - GR,O,UP
@STRUMENTS. i i.
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- 6.3.6 THE CONTROL ROOM CONYENTION PROJECT CONTAINS 12 HEDS.
WHICH ARE COVERED IN 13 PLANT MODIFICATIONS. THE CONTROL ROOM CONVENTION PROJECT SURVEYED ALL ANNUNCIATORS' TILE ENGRAVING. PANFL LABELING. COMPONENT LABELING. FUNCTION LABELING, AND POSITION LABELING.
THESE ITENS WEAE COMPARED TO IDENTIFY INCONSISTENCIES AND INCORRECT USAGE WITH ABBREVIATIONS. THE SURVEY ALSD ADDRESSED THE APPLICATION OF COLOR CODING IN THE CONTROL ROOM AND DEDICATED SHUTDOWN PANELS AND CONTROL DIRECTIONAL MOVEMENT.
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. I 6.3.7 THE CONTROL ROOM HVAC PROJECT CONTAINS ONE HED, WHICH IS COVERED IN ONE PLANT MODIFICATION. THE HVAC SURVEY ADDRESSED THE TEMPERATURr AND HUMIDITY LEVELS IN THE ;
C0pTROL ROOM, HOT / COLD SPOTS DRAFTS, RELIABILITY, AND OPERATOR COMMENTS.
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6.3.8 T9E OFF-GAS FLOW INSTRUMENT PROJECT CONTAINS ONE HED
- WHICH IS C0VERED BY ONE PLANT MODIFICATION. THE :
OFF-GAS DUTLET FLOW RECORDER IS FREQUENTLY OUT OF SERYlCE AND IS NOT RELIABLE. ,
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- I ENGINEERING STATUS OF HED PROJECTS: i
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. . . .- e, F i l 6.3.3 THE COMP 0NENT REM 0 VAL PROJECT 1 l
COMM11NENT: UNIT 1 REFUELINs OuTAsE 8 (1992) ,
UNIT 2 REFUELING OuTAsE 10 (1993) I PLANT MODIFICATION STATUS: -
j Six PLANT MODIFICATIONS HAVE SEEN COMPLETED.
FIVE PLANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION
! IN,1989.
j TWO PLANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION IN 1990.
)
TWO PLANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION IN l
1991. - i FIVE ?LANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION .
IN 1992, e
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UNIT 2 REFUELING OUTAGE 9 (1991)
PLANT MODIFICATION STATUS: -
THREE PLANT MODIFICATIONS HAVE BEEN COMPLETED.
l- ONE PLANT MODIFICATION IS SCHEDULED FOR COMPLETION IN 19,89.
TWO PLANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION IN 1990.
ONE PLANT MODIFICATION IS SCHEDULED FOR COMPLETION IN 1991. -
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6,3.5 TNE CoMroNENT RELOCATION PROJECT l
CTellTNENT: UNrT ! REFUELING OuTAsE 8 (1992)
, UNIT 2 REFUELING OUTAGE 10 (1993) l PLANT MODIFICATION STATUS:
l
. 1 FouR PLANT MODIFICATIONS HAVE sEEN COMPLETED.
THREE PLANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION IN,1990. :
I SIXTEEN PLANT MODIFICATIONS ARE SCHEDULED FOR i COMPLETION IN 1991.
EIGHT PLANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION IN 1992.
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6.3.6 I TNE CONTROL Ro0M CONVENTION PROJECT l
i CIMMITHENT: UNIT 1 REFUELING OUTAGE 8 (1992)
, UNIT 2 REFUEllHG OUTAGE 10 (1993)
PLANT NODIFICATION STATUS: -
TWO PLANT MODIFICATIONS HAVE BE'J~. COMPLETED.
TWO PLANT N0DIFICAT!0NS ARE SCHEDULEL, FOR COMPLETION IN 19$9.
THREE PLANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION IN 1990.
FIVE PLANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION IN 1991.
- ONE PLANT MODIFICATION IS SCHEDULED FOR COMPLETION IN 1992, i i
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THE CONTROL ROOM HVAC PROJECT CCMMITMENT: UNIT 1 REFUELING OuTAst 7 (1990)
. UF'T 2 REFUELING OUTAGE 7 (1990)
Pl. ANT MODIFICATION STATUS: -
THis PLANT MODIFICATION IS SCHEDULED FOR COMPLETION IN l 1990.
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t 6.3.8 THE OFF-6As Flow INsinuMENT PROJECT C3MITNENT: UNIT 1 REFUELING OUTAGE 8 (1992) !
UNIT 2 REFUELING OUTAGE 11 (1993) . l PLANT MODIFICATION STATUS: -
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THESE PLANT N0DIFitAT10N$ ARE SCHEDULED FOR COMPLETION IN 1989. ;
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TOTAL PLANT MODIFICATION STATUS:
FirTEEN PLANr MODIFICATIONS HAVE BEEN COMPLETED.
1 EIGNT PLANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION IN 1989.
l TEN PLANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION IN 1990. j l
TWENTY-F0UR PLANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION IN 1991.
FOURTEEN PLANT MODIFICATIONS ARE SCHEDULED FOR COMPLETION IN 1992.
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, s 6.2.10 control R,oom y intenance Project - To be completed by December 44r,'-tMS, i
MEDs to be addressed: 15 Total
@ A.4 5 l 2263-3259/U 20J7-2390 (d 206X-1189(M
/, ,. A. .r . ' . 21X5-3760 ( 206X-2414Q) 206X-1192 (F) 206X-3581 (@ 20X4-2487 (5) 20F1-1939 (6) i 2063-2227 (C 206X-1122U) 216X,0119Gd ,
l 206X-2349 (s) 206X-1125(d 404X-21288G. h rol Room Furnishings Project -
To be plated by Deceiiibe 1987.
REDS to be addr i 15 Tota
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20sY-3204 2103-0103 20H1-0108 2.12 Training Project - To be completed by. December 31, 1987.
MEDrto addressed: 8 Total ,
l 20N0-0501 OX 4 2001-0315*
2050-0506 001-0306 0 -
5012 20X5pH 58 2001-0312 6.2.13 g Coding Project - TobecompletedbyDecemberIb ,
Jr988'. , ,
MEDs to be addressed: 7 Total l }
d'- O*% 3 2063-2202 &) 206X-2241(9 2063-1126 ( 9 {
A. /4J., 5 206X-2209(5) 206X-2245 (5) 20X5-5015*(p i 20X1-2240 (s)
- EED is addressed in more than one project. REV. 1 6-5
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- e. M t ek Radiation Monitor Recorde 'nstallation
, , Project - been completed nit 1. To be ,
completed by t end --R ueling Outage G for Unit 2 L
(currently sc 01/02/88 to 04/22/88). !
HEDsj addressed: Total ,
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i 6.3 Newly Scheduled Proiects
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The following projects have been evaluated to determine the appropriate corrective actions for the MEDs listed and , i implementation scheduled developed. !
t 6.3.1 Annunciator Project - To be completed by the end of ;
Refueling Outage 4 for Unit 1 (cur?lently scheduled [
- - /
-$ 3 02/15/92 to 05/08/bt) and Refueling odtage 9 for Unit 2 !
' * ' "- % (currently schedu.ed 04/27/91 to 07/19/91). l NEDO to be addrsssed: 10 Total N I 20N0-2102* 4 0 206X-2116 W)" 206X-2124 u- Al-706X-2106 #C C 206X-2117uff' -106X-21M Ir$
20H0-H44 h 206X-2120 09' 20X3-2129td" f a c..ai206X-2115 49 ', l 6.3.2 eof Instrumention Project - To be completed by the end !
of Refueling outage 7 for Unit
- 1 (currently scheduled J / e'f4 '
07/07/90 to 10/05/90) and Refueling outage 9 for Unit 2 (currently scheduled 04/27/91 to 07/19/91). ,
HEDs.to be addressed: 6 Total ,
- i -206X-5096(Oca d ,
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, 206X-5092 @ '* -206X-50970) M /
206X-5094Ofa 206X-5093 (C" t
- EED is addressed in more than one project. REV. 1 6-7
--u - A L.....
. S 6.3.3 Component Removal Project - To be completed by the end of Refueling Outage 8 for Unit 1 (currently schedulbd 02/15/92 to 05/04/92) nnd Refueling outage 10 for Unit 2 (currently scheduled 11/28/92 to 02/26/93).
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MEDg to be addressed: 1 Total 2061'-Th)33(M 4 A4 3 .
/ . rM f 206X-3003 (#' 206X-1409*$D -204X-2222(O* tu '
y . / wit 20X5-5004 W' 206X-3255 (.D" MJ2-1916 (5?"*/"
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20X3-5003 M ' 20X2-5083LO" 20X3-5068*N)"
20X3-5032U7 43J2-1416(U 20F1-5088C.0" 20X3-5071 # ' 24df-1414th 2061-14'04(I) "
6.3.4 Indicator Upgrade Project - To be completed by the end of Refueling outage 8 for Unit 1 (currently scheduled 02/15/92 to 05/08/921 and Refueling outage 9 fer Unit 2 (c'trrently scheduled 04/27/91 to 07/19/91). ,
! HEDs to be addressed: 16 Total
- S /^*% 2 . 20X3-2413(27 206X-2226dY 20X5-2406*@"
l 1 A-4 3 2163-2416(5) " 206X-1146(J[ 206X-2230 (d
p . Ts.dtS 206X-1145U7 206X-1187(C" 206X-2479(0" 206X-2228U)" 2063-5019(.V 204M-3444+0 ~4
! 206X-2080(C" 20X5-240$*(r)[ 20X8-5021 (Q
20$X-2225(s) .
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- HED is addressed in more than one project. REY. 1 6-8 -
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6.3.5 Component Relocation Project - To be completed by the end '
. of Refueling Outage 8 for Unit 1 (currently scheduled i f f l' 02/15/92 to 05/08/92) and Refueling outage 101'or Unit 2
,(currently scheduled 11/28/92 to 02/26/93).
.
- MI:Ds to be addressed: 32 Total
/TQ 20X2-1162 M 206X-5069(d 20X2-1169CM 20X3-507007 20X5-5079 I v /d 3 20X5-5080(f[
- 5. raJ 5 '
20X2-5085 d 206X-5072(C" 2061-5043 (d 206X-5036D7 206X-5073(C" 20X2-2201 ([
206X-50510*? 206X-507.6(If 206X-1104 (#
! 20X2-5052 # 206X-5081(D' 206X-1113 (tf I 20X2-5053(D" 206X-5033 UY 206X-1151 8 I 206X-506007 206X-5057U7 2063-1170 (
206X-5064(>f 20X5-5061/8 34Na-3243*44) l
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206X-5063([ * '
20X2-5065 # 20X2-5062 0)" I 20X3-5068*(d 206X-5078 UI l !
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6.3.6 Control Room convention Freject - To be completed by the ;
and of Refueling Outage 8 for Unit 1 (currently scheduled ;
/ /'H 1 02/15/92 to 05/08/92) and Refueling outage 10 for Unit 2 jo , /1f 3 (currently scheduled 11/28/92 to 02/26/93).
i - fd 5 EEDs to be addressed: 12 Total 20RS-3289 Of' 20X2-1302(9 " 20X5-3218(D "
20X5-1105*(W/ 206X-1304(d" 20X2-3268 (0 206X-1163 (U" 2063-1305(3)" 206X-5038G)"
206X-1178 (8) 206X-1306 (J)" 206X-5082 (D 1 .
- 6.3.7 Control Room NVAC Project - To be completed by the end of
. . Refueling Outage 7 for Unit 1 (currently scheduled 07/07/90 to 10/05/90) and Refueling Outage 8 for Unit 2 (currently i scheduled 09/21/89 to 11/01/89).
EED to be addressed: 1 Total 20n0-0002(3)"
- KED is addressed in more than one project. REV. 1 6-9
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.. 6.3.8 offgas Flow Instruments Project - To be completed by the and of Refueling outage 8 for Unit 1 (currently scheduled l I" 3 02/15/92 to 05/08/92) and Refueling outage 10 for Unit 2 (currently scheduled 11/28/92 to 02/26/93).
MED to be addressed: 1 Total
.'. 20X8-5018 (9"
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6.4 Additional Corqmitments ,
6.4.1 Development and implementation of a Human Factors Design Guide Project - To be completed by December 31, 1987. l 6.4.2 ERFIS and SPDS Survey - This project was a commita nt l l
made in the CRDR program plan that could not be completed i
! by the Final Summary Report submittal date. This will be I completed as part of the ERTIS/SPDS project, within 3 months after Refueling Outage 6 for Unit 1 (currently scheduled 10/15/88 to 01/06/89) and within 3 months after Refueling Outage 7 for Unit 2. !
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