ML20245G760
| ML20245G760 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 04/12/1989 |
| From: | SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| Shared Package | |
| ML20245G740 | List: |
| References | |
| CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-89-1121, TAC-M59329, NUDOCS 8905030226 | |
| Download: ML20245G760 (25) | |
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SAIC-89/1121 l
IN-PROGRESS AUDIT REPORT OF THE DETAILED CONTROL ROOM DESIGN REVIEW AT BOSTON EDISON COMPANY'S PILGRIM NUC'. EAR POWER STATION TAC NO. M59329
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April 12, 1989 Prepared for:
U.S. Nuclear Regulatory Comission Washington, D.C. 20555 Contract NRC-03-87-029 Task Order No. 35 r
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TABLE OF CONTENTS Section E121
1.0 INTRODUCTION
I 1.1 Background.................................................
1.2 March 1989 Audit Agenda and Participants...................
2 2.0 EVALUATION....................................................
3 2.1 Establishment of a Qualified Multidisciplinary Review Team.......................................................
3 2.2 System Function and Task Analysis..........................
4 2.3 Comparison of Display and Control Requirements with a C o nt rol Ro om I nv en t o ry.....................................
6 2.4 C o n t rol Ro om S u rv ey........................................
7 2.5 Assessment of Human Engineering Discrepancies (HEDs) To Determine Which Are Significant and Should Be Corrected....
8 2.6 Sel ecti on of Design improvements...........................
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l 2.7 Verification that Selected Design Improvements Will Provide the Necessary Correction...................................
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2.8 Verification the Improvements Will Not Introduce New HEDs..
12 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 operating Procedures................
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3.0 CONCLUSION
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4.0 REFERENCES
If List of Audit Participants Observations by Audit Team l
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IN-PROGRESS REPORT FOR THE DETAILED CONTROL ROOM DESIGN REVIEW AT BOSTON EDISON COMPANY'S PILGRIM NUCLEAR POWER STATION MARCH 20-21, 1989
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1.0 INTRODUCTION
This report documents the findings of an in-progress audit of the Boston Edison Company's Detailed Control Room Design Review (DCRDR) program at Pilgrim Nuclear Power Station. The audit was conducted by the Nuclear Regulatory Commission (NRC) during a site visit the week of March 20, 1989.
The purposes of the audit were:
o To assess the licensee's progress toward completing the DCRDR requirements stated in FUREG-0737, Supplement 1 (Reference 1) since the previous DCRDR andit conducted in November 1984.
o To discuss the licenis a's plans and projected schedules for completing the OCRDR praynm, with the goal of ensuring that all requirements will be met 21 the earliest possible date.
To achieve a mutually clear understanding between the licensee and o
the NRC staff about the information to be provided in upcoming licensea submittals -- i.e., the program plan revision and summary report supplement which are scheduled for receipt by the NRC no later than April 30, 1989.
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Background
The following is a chronological list of major milestones to date in the Pilgrim DCRDR.
10/83:
Program plan for conducting the DCRDR submitted by licensee (Reference 2).
184_;.
NRC staff comments on program plan issued (Reference 3).
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3/El Revised program plan submitted by licensee (Reference 4).
9/84: Sumary report on the DCRDR submitted by licensee (Reference 5).
11/84:
Preimplementation audit of DCRDR findings conducted by NRC staff.
1/El Safety evaluation report (SER) issued by the NRC, documenting open items and requesting date for submittal of a supplemental sumary report on the completion of those items (Reference 6).
The preimplementation audit, as documented in the SER (Reference 6),
determined that additional work was needed to complete eight of the nine DCRDR requirements of NUREG-0737, Supplement 1.
The licensee was requested
- to complete this work and submit a supplemental sumary report describing methods and results. The work has not been completed as of the date of this audit.
The licensee has comitted to submit, by April 30, 1989, a supplemental sumiry report that describes what has been accomplished since the SER was issued.
The licensee has also committed to submit a
supplemental program plan to describe the processa that will be used to complete the work, with schedule projections indicating when the remaining tasks are expected to be completed.
1.2 March 1989 Audit Agtnda and Participants
'The licensee provided an opening sumary of program status -- work in progress, work to be done, and post-SER accomplishments. Thereafter, each of the nine DCRDR requirements of NUREG-0737, Supplement 1, were reviewed in detail, by means of discussion sessions and observations in the control room and simulator. A technical discussion of the audit findings was conduct 3d with the licensee's DCRDR project team.
In addition, the findirgs were sumarized in a formal exit briefing given by the NRC audit team leader.
The audit was conducted by a team of NRC staff and contractors from Science Applications Internation:1 Corporation (SAIC), and a member of the Spanish Nuclear Regulatory Agency, representing the disciplines of human factors engineering, nuclear operations, and instrumentation and control 2
systems engineering.
The licensee team included members from several divisions within the Nuclear Engineering Department, which has responsibility for all NUREG-0737, Supplement 1 initiatives, as well as representatives from the OperTtions Department,. and human factors engineering consultants.
See Attan iaent I for a list of participants.
2.0 EVALUATION The status of the Pilgrim Nuclear Power Station DCRDR is evaluated with
. respect to each of the nine requirements stated in NUREC 0737, Supplement 1, in the following sections.
2.1 Establishment of a Qualified Multidisciplinary Review Team The organization for conducting a successful DCRDR can vary widely, but is expected to conform to some general criteria.
Overall administrative leadership should be provided by a utility employee, who should be given 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 for the technical l
l tasks.
Finally, the DCRDR team should receive an orientation briefing on J
DCRDR purpose and objectives which contributes to the success of the DCRDR.
l, NUREG-0800, Section 18.1, Appendix A (Reference 7) describes criteria for the niultidisciplinary review team in more detail.
l It was previously determined (SER, Reference 6) that the licensee established an appropriate administrative organization and a qualified multidisciplinary team to conduct the DCRDR. The March 1989 NRC audit determined that the administrative structure and the multidisciplinary
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composition of the licensee's DCRDR team have been maintained, with several
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of the original personnel still involved.
The use of operator input in the development and verification of control room enhancements was noted as a highly valuable practice. The licensee indicated the intention to continue this practice in developing and impleme ing all HED corrective actions.
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i The licensee's team composition for completing the DCRDR was not final at the time the audit was conducted.
This NUREG-0737, Supplement 1,
requirement remains to be completed. The licensee stated that staffing will be described in the program plan revision scheduled for submittal to the NRC by April 30, 1989.
It is understood that the team composition for each remaining task will be defined in terms of disciplines, as a minimum.
Specific team members will be identified insofar as possible.
2.2 System Function and Task Analysis The purpose of the system function and trsk analysis (hereafter referred to as 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:
1.
Analyze the functions performed by systems in responding to transients and accidents in order to identify and describe those i
tasks operators are expected to perform.
2.
Determine what information (e.g.,
parameter,
- value, status) i l
signals the need to perform the task, the control capabilities needed to perform the task, and the feedback information needed to monitor task performance for each task identified in Item 1 above.
3.
Analyze the information and control capability needs identified in Item 2 above to determine appropriate characteristics for displays and control to satisfy those needs.
The licensee has decided to completely revise the DCRDR task analysis for two major reasons. The SER (Reference 6) identified deficiencies in the method used to perform the original task analysis.
It was determined that the licensee did not perform a systematic front-end analysis to establish operator information and control requirements, including the specific instrumentation and contisi characteristics necessary for operator tasks.
l Instrumentation and control requirements were not analyzed independently of the existing control panels and the analysis did not extend to the level of 4
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i necessary detail; i.e., it did not specify characteristics such as scale range, units of measure and resolution needs for trend versus point data, type and precision of ;vntrol action.
Without documentation of an analytically established set of requirements, the suitability of the existing controls and displays could not be evaluated adequately.
In addition, the crevious analysis was based on plant-specific emergency operating prn.3dures (EOPs) deveioped from Revision 2 of the Boiling Water Reactor (BWR) Owners Group generic Emergency Procedure Guidelines (EPGs). The NRC has approved Revision 3 of the generic EPGs as the minimum acceptable functional requirements baseline for the DCRDR task analysis (Reference 9). A major difference in Revision 3 of the generic EPGs is the sddition of guidelines for secondary containment control and radioactivity release control. The licensee's original task analysis did not address these emergency response requirements.
The licensee has updated the Pilgrim Nuclear Power Station E0Ps to reflect the guidance in Revision 3 and Revision 4 of the generic EPGs.
The licensee stated that the new task analysis will reflect the changes introduced by both Revision 3 and 4 of the EPGs.
Discussion of the task analysis during the March 1989 audit indicated that the responsible personnel on the current Pilgrim DCRDR team have a clear understanding of the methodological deficiencies of the previous task analysis and of the elements of a useful task ar.alysis. The licensee stated thatIthe new task analysis approach will include the following:
o Performance by a multidisciplinary team including operations, engineering, and human factors.
Analysis of all nine plant-cpecific symptom-ba:ed E0Ps, current to o
Revision 4 of the EWR Owners Group EPGs.
o Analysis of satellite procedures referenced in the E0Ps, insofar as applicable to emergency situations.
o Listing of all operator action requirements reflected in entry conditions, steps, cautions, and notes.
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Listing of controls and
- displays, and their specific characteristics, from analysis of the operator action requirements.
The licensee stated that the task analysis method and staffing will be described in detail in the next program plan revision.
Since the task analysis is to be redone, this NUREG-0737, Supplement 1, requirement remains incomplete. The licensee was requested to provide a schedule for the task analysis and related activities in the next program plan revision, including start, milestone, and completion dates.
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 by operators to perform their tasks. The success of this element depends on the quality of the 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 results cf the function and task analysis.
Unavailable or unsuitable displays and control should be documented as human engineering discrepancies (HEDs).
'A cot 1puter-based control room inventory was created during the original DCRDR effort at Pilgrim. The licensee plans to update this inventory to reflect changes made to the control panels. The licensee plans to use the computer-based inventory for comparison to the information and control requirements defined in the task analysis.
I The value of conducting this comparison using the control room itself, L
in the context of E0P walk-throughs, was also discussed.
The audit included walk-throughs of two E0P samples to illustrate this method.
The method included a control room operator, a human factors specialist, and an engineering specialist to walk through the E0Ps, discussing each action I
requirement and the displays and controls requirements to perform the actions.
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l A number of potential human engineering discrepancies (NEDs) were identified during the audit walk-throughs. They are listed in Attachment 2 to this report. This list is provided for use by the licensee in evaluating the completeness of discrepancy identification in DCRDR activities to date.
The licensee indicated that many of the items had been previously
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identified, and provided a partial list of corresponding HEDs. The licensee should ensure that all of the items listed in Attachment 2 are evaluated during the processes of completing the DCRDR, and that they are added to the set of HEDs as appropriate.
I The comparison of control and display requirements to the control room inventory will be redone.
Therefore this NUREG-0737, Supplement I
requirement remains incomplete. The licensee was requested to describe the
- method, staffing, and schedule for conducting this comparison in the upcoming program plan revision.
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 Section 6 of NUREG-0700 (Reference 8); however, other accepted human factors standards may be chosen. Discrepancies should be documented as HEDs.
The licensee previously performed portions of the control room survey using checklists developed from human engineering criteria in NUREG-0700, the BWR Owners Group control room survey checkb a, and Nuclear Utility Task Action Committee (NUTAC) control room survy documents.
The SER concluded that the survey methodology and checklist: ure adequate except that three NUREG-0700 criteria were not, and should be, addressed (see Reference 6, p.13).
The NUREG-0737, Supplement 1, requirement for a control room survey has not been met because the survey is incomplete in the following respects:
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o The computer survey has not been done. Workspace layout criteria pertinent to use of the computer in the control room have not been applied.
o The HVAC survey has not been completed.
o The noise survey has not been completed.
o Parts of the control-display integration and panel layout surveys have not been completed.
In addition, changes have been made in the control room since the previoLs survey work was performed. Many of these changes originated outside of the DCRDR and were not implemented under the control of the DCRDR team. The licensee stated the intention to update the previous survey work to cover changes made since that time.
The licensee should specify the
- staffing, method (including checklists),
and schedule for performing the remaining survey tasks in the upcoming program plan revision.
2.5 Assessment of Human Engineering Discrepancies (HEDs) to Determine Which Are Significant and Should Be Corrected Based on the guidance of NOREG-0700 and the requirements of NUREG-0737, Supplement 1,
all HEDs should be assessed for safety significance.
The potential for operator error and the consequence of that error in terms of plant safety should be systematically considered in the assessment.
Both the individual and aggregate effects of HEDs should be considered.
The result of the assessment process is a determination of which HEDs should be 1
corrected because.of their potential impact on plant safety. Decisions on whether HEDs are safety-significant should not be compromised by consideration of such issues as the means and potential costs of correcting HEDs.
The assessment process originally used to determine which HEDs should be corrected was found in the SER (Reference 6, page 18) to meet the assessment requirement of NUREG-0737, Supplement 1.
However, the licensee l
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decided to develop a new assessment process and to reassess the previously identified HEDs. The reassessment has not been completed.
The licenne stated that the new assessment process has beet, formalized in a Nuclear Engineering Department Working Instruction, NEDWI-392.
The new method considers risk and consequences of error expressed in terms of an index of error likelihood and and index of costs of error consequences.
It was not clear whether or how these two indexes will be combined to achieve a single prioritization 'of HEDs according to safety significance.
The licensee indicated that there will be no rank cutoff for deciding which HEDs should be fixed; the decision will be made on a case by case basis using the assessment results as decision input.
The licensee should provide a complete description of the ne'w kssessment method in the upcoming program plan submittal. The description should include identification of the team of personnel (disciplines) who performed the reassessment of old HEDs and who will be responsible for assessment of new HEDs identified in future DCRDR activities.
The description should provide assurance that safety considerations, not the cost of corrective actions, will drive decisions about which HEDs should be l
corrected, Since the assessment is incomplete, this requirement of NUREG-0737, Supplement I remains open.
2.6 Selection of Design Improvements The purpose of selecting design improvements is to provide means to correct HEDs identified from the review phase of the DCROR.
Selection of design improvements should include a systematic process for the development and comparison of alternative means of resolving HEDs.
Furthermore, according to NUREG-D737, Supplement 1, the licensee should document all of the proposed control room changes.
The SER (Reference 6) determined that the methodology for selecting design improvements, as described in the revised program plan submitted in August 1984 (Reference 4) appeared adequate for successful completion of this requirement of NUREG-0737, Supplement 1.
In the upcoming program plan i
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revision, the lice.
,e should describe any changes to the previously defined processes and responsibilities for selecting and developing corrective j
f actions.
The licensee stated at the preimplementation audit, in November 1984,
" Design Manual" would be prepared to help standardize the Pilgrim that a control room modifications.
It was stated that the manual would be usH to ensure consistent and compatible HED solutions. The intent of a c< sign
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manual has been partly achieved by the preparation of three engineering l
standards:
E-543, Design of Nameplates and Labels o
i E-544, Design of Instrument Scales o
E-545, Design of Control Panel Demarcation.
o Minimal progress has been made in selecting, developing, and finalizing corrective action plans since the SER was issued.
For example, of the eight HEDs classified as safety significant (Category "A")
by the original assessment process, three have been corrected (HED Nos. 4A003, 5A004, and 5A005).
The audit team found that those corrective actions were appropriate.
The licensee stated that a design change request to correct of the Category A HEDs (No. 8A006) has been approved; this will one more involve surface enhancement of four panels. No decisions have been made about whether, how, or when to correct the remaining four HEDs that were determined to be safety significant in the original assessment.
There are 145 other HEDs originally assessed as Category B or C (B
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associated with safety considerations; C - associated with availability or reliability considerations). The licensee stated that approximately 20 of these HEDs have been corrected.
In the upcoming supplemental summary report, the licensee should l
I identify the correction status of each HED documented in the previously submitted summary report (Referenca 5). The ifcensee should also provide descriptions of any new HEDs identified since the original summary report was submitted, and should include those new HEDs in the summary of HED correction status.
The licensee should classify the HEDs according to status category; for example: correction implemented (briefly describe the 10 l
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nature of the correction); correction scheduled for implementation (provide C
date/ time frame); correction designed or in design; decision made, with justification, nol to correct HED; no decision about HED correction.
Since the selection of HED corrective actions is incomplete, this l
requirement of NUREG-0737, Supplement 1, remains open.
2.7 Verification that Selected Design Improvements Will. Provide the Necessary Correction A key criterion of DCRDR success is a consistent, coherent, and effective interface between the operator and the control room.
This of criterion may be met by effectively executing the processes of selection will provide design ' improvements, verification that selected improvements will not the necessary correction, and verification that the improvements introduce new HEDs.
According to NUREG-0800, techniques for the i
verification process might include resurveys of panels, applied experiments, engineering analyses, environmental surveys, and operator interviews.
The consistency, coherence, and effectiveness of the entire operator-control interface are important to operator performance. Thus, evaluation of room both the changed and unchanged portions of the control room is necessary during the verification process.
The licensee has not specifically described the process of verifying design improvements in previous DCRDR submittals. For surface enhancements introduced to date, the licensee is using a process of trial implementation the simulator and review / revision in that setting based on input by the in control room operators. There is human factors input to this process since the' surface enhancements are being implemented by a human factors contractor.
This provides an opportunity for verification of the The effectiveness of design improvements before their final installation.
licensee should devise equivalent strategies for verifying design improvements, other than surface enhancements, and for addressing the selection and evaluation of longer-term design changes.
For final, post-implementation verification, the licensee has specified requirements and forms in a Nuclear Engineering Department Working Instruction (NEDWI-392). The licensee has also established a job assignment 11
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for tracking HED corrective actions to ensure that they are properly completed and closed out. The licensee should summarize these and other features of the design improvement verification process and responsibilities in the upcoming program plan revision.
A great deal of work remains in developing, and implementing design improvements to correct HEDs. Therefore, the adequacy of the licensee's processes for verifying the effectiveness of design improvements cannot be fully evaluated at this time. This verification requirement of NUREG-0737, Supplement 1, discussed in this section, remains open.
2.8 Verification the Improvements Will Not Introduce New HEDs As discussed in Section 2.7 above, the licensee does not have a formal process for verifying that the proposed improvements will not introduce new HEDs when implemented. Therefore, this verification requirement of NUREG-0737, Supplement 1, remains open.
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 Operating Procedures Improvement of emergency response capability requires coordination of the.DCRDR with other activities.
Satisfaction of Regulatory Guide 1.97 requirements and the addition of the Safety Parameter Display System (SPDS) necessitate modifications and additions to the control rtnm.
The modifications and additions should be specifically addressed by the DCRDR.
Exactly how the modifications are addressed depends a number of factors including the relative timing of the various 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 l
a consistent,
- coherent, and effective control room interface with the operators.
I The SER (Reference 6) documented that the licensee had defined l
appropriate methods to coordinate the several programs required by NUREG-l 0737, Supplement 1, to improve emergency response. The SER noted,
- however, I
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that coordination between the E0P upgrade program and the DCRDR needed improvement.
During the March 1989 audit, other needs to close coordination loops were identified. Regulatory Guide 1.97 instrumentation on the panels has not been identified, and new 1.97 instrumentation added to the panels has not been reviewed for human factors engineering adequacy.
As another example, consistency between parameter values in SPDS displays and the l
related panel instrument readings has not yet been verified.
l In presentation and discussions during the March 1989 audit, the licensee described a number of useful mechanisms planned to achieve coordination of other programs with the DCRDR.
These plans have not been formally documented and submitted to the NRC.
The licensee's entire coordination strategy should be summarized in the upcoming program plan revision.
The effectiveness of coordination cannot be evaluated until the plans are implemented.
Therefore, this requirement of NUREG-0737, Supplement 1, remains open.
3.0 CONCLUSION
The basic conclusion of the March 1989 audit is that minimal progress has been made toward completing DCRDR requirements since the fall of 1984, when the original summary report was submitted and a preimplementation audit was conducted by the NRC. Substantial additional work remains to complete the ' processes of HED identification.
Of the 153 HEDs identified in the original summary report, the licensee stated that only approximately 24 have been corrected. Decisions about corrective actions have not been made for the majority of the remaining HEDs in that original set.
The potential exists, also, for identifying additional HEDs as a result of implementing a new task analysis, revising the survey, and integrating the DCRDR with changes from other programs.
All nine requirements stated in NUREG-0737, Supplement 1,
for the performance of a DCRDR remain open.
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4.0 REFERENCES
l 1.
U.S. Nuclear Regulatory Comission (1982). Requirements for emergency response capability.
NUREG-0737, Supplement 1, issued by Generic Letter 82-33.
2.
Boston Edison Company (1983). Detailed control room design review program plan.
Attachment I to letter from W.D.
Harrington, Boston Edison Company, to Domenic b. Vassallo, NRC, October 14, 1983.
3.
U.S.
Nuclear Regulatory Comission (1984). Staff coments on the Pilgrim Station I detailed control room design review.
4.
Boston Edison Company (June 1984). Detailed control room design review I
program plan. BEC084-134, Revision 1, forwarded by letter from W.D.
Harrington, Bedon Edison Company, to Domenic B. Vassallo, iE, August l
14, 1984.
j S.
Boston Edison Company (September 1984). Detailed control room design review executive sumary report.
CEC 0/ESR-1, Revision 1.
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6.
U.S.
Nuclear Regulatory Comission (1985).
Safety evaluation by the Office of Nuclear Reactor Regulation of the detailed control room.
design review for Pilgrim Nuclear Power Station, Docket No. 50-293.
i 7.
IV.S.
Nuclear Regulatory Comission (1984).
Standard review plan' of safety analysis reports for nuclear power plants, Appendix A to Section 18.1, control room design review.
NUREG-0800, Revision 0.
1 8.
U.S. Nuclear Regulatory Comission (1981). Guidelines for control room design reviews. NUREG-0700.
9.
U.S. Nuclear Regulatory Comission (May,1984). Meeting Sumary - Task Analysis Requirements of Supplement 1 to NUREG-0737 - May 4,
1984 Meeting with BWR Owners Group Emergency Procedure Guidelines and Contrcl Room Design Review Comittees.
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i ATTACHMENT 2 LIST OF PARTICIPANTS c
LIST OF PARTICIPANTS NRC Audit Team James P. Bongarra, Jr., Office of Nuclear Reactor Regulation (NRR)
Richard P. Correia, NRR Rafael Cid, Spanish Nuclear Regulatory Agency Gordon R. Bryan, Jr., COMEX Corporation Barbara Paramore, Science Applications International Corporation NRC Proiect Manaaer 1
Dan Mcdonald Boston Edison Comoany and Contractors Marie Lenhart, Licensing, audit coordinator V. 0heim, Nuclear Engineering Department (NED), Deputy Department Manager David Bryant, NEO, DCRDR Project Manager C.H. Minott, NED, Deputy Project Manager Warren Babcock, Jr., NED, Control Systems Design, DCRDR Principal Investigator Siben Dasgupta, Control Systems Division Manager Norm Eisenmann, Control Systems Division, HED tracking Steve Brennion, NED, System Safety Analysis Division (SASA) E0P dev. manapr Kathy Ward, SASA, HED assessment methodology L. Olivier, Operations Department, Chief Operating Engineer Ken Taylor, Operations Department (SRO), DCRDR liaison Danna Beith, Management Analysis Corporation (MAC), human factors Rett Considine, MAC, human factors
-s ATTACHMENT t OBSERVATIONS BY AUDIT TEAM DURING CONTROL ROOM WALKTHROUGHS 4
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ATTACHMENT 2,-
q OBSERVATIONS BY AUDIT TEAM-DURING CONTROL ROOM WALKTHROUGHS l.
Labeling of recorders is incomplete:
Examoles:
o Hydrogen concentration recorders on PAM panel -- parameters traced by pens are not identified.
t o
Labels missing on drywell temperature / pressure recorder, Panel 903, left end.
o Labels missing on two recorders below drywell mimic on Panel 903.
o Recorder CRU 3361 on Panel Cl is unlabeled as to ID and engineering units.
2.
Recorder pen color assignments are inconsistent:
Examoles:
.o Recorders 640-26 and 640-27 on Panel 905 -- Reactor Steam Flow is traced by the red pen on one of these recorders and by the blue pen on another, o
A similar inconsistency exists on the wide range and narrow range pressure recorders.
3.
The SCRAM solenoid indicator lights are on a back panel behind the main control panel (horseshoe).
It is necessary to check these lights to verify that SCRM has occurred (that the plant is not in an ATWS condition). These four lights should be up front on Panel 905.
4.
The wording of Caution 1, in E0P-01, is confusing. However, it appears to require the operator to determine the usability of RPV water level 1
i instrumentation in the control room by one of two methods:
(1) comparing temperatures near the instrument reference leg vertical runs to a criterion value, " maximum RB run temperature"; or (2) comparing i
the control room instrument readings to the criterion value of " minimum f
usable l evel. "
To use the temperature criterion, the control room operator would have to send an A0 to take readings in the pl ant.
1 Regarding the level criterion, there is insufficient scale range on the Fuel Zone level indicator and recorder to determine in all cases whether level is above the specified minimum of -263 in.
l 5.
E0P-01, Figure 1.1, requires indication of Torus Pressure in the high j
range.
There is none in the control room. The PAM panel Containment l
Pressure High meter, PI-1001-600B, is used (incorrectly) to provide this parameter.
l 6.
To determine torus pressure >2.5 psig, the operator must perform a calculation using drywell pressure and differential pressure.
7.
The scale faces on PID-5067A and 5067-8 are incorrectly labeled psid.
The correct unit of measure is psig.
8.
E0P-01, Caution 2, requires the operator to ensure HPCI turbine speed
>715 rpm.
The meter scale is in increments of 50 rpm.
It is also missing a label.
J 9.
The HPCI vibration meter on Panel 903 is an " abandoned" component and is not so labeled. The same is true of the N2 recycle blower isolation i
valves.
(These and other abandoned components should be removed.)
10.
E0P 0.1 specifies top of active fuel (TAF) as
-126.3.
An engraved an obsolete Panel 903 gives TAF as -127.5 inches, which is label on l
value.
- 11. There are numerous problems with labels and nomenclature:
l Some meters still carry construction dot marks.
o o
PP vs pump j
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- O o
Inbd vs inboard o
V1v vs valve o
SUPP vs suppression vs torus o
The terms torus, suppression chamber, and containment seem to be used interchangeably.
Spelling (ampers on SW, feed, and condensate pumps; RESIVOIR on C-o 2) o Dp vs A P vs delta P o
Black on white and white on black o
H202 on PAM panel right, B-6 o
SRM HI/INPO vice SRM HI/INOP on 905R, B2 o
Missing label on FC 5030B flow controller o
Tape peeling from the ADS inhibit label
,o Paint-on labels:
containment spray full flow test; B core spray shutdown with auto start locked in; #4 control valve above seat drain
- 12. Mimics are not yet provided where they would be helpful. Also, panel layout discrepancies were obr.erved:
o No mimics for HPCI, RHR (all modes),
core spray, MSIVs and bypass / drain valves, RCIC, RWCU, feed and condensate demin (except for backboard heater mimic), CRD drive and cooling, recire loops, RBCCW, steam seal and SJAEs, and TBCCW o
Steam line drains /MSIV bypasses are separated from the MSIVs by RCIC.
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RWCU valves are separated from head vents and DW/ torus sample controls.
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o Feedwater heater controls are located on back panel C4, separate from the other feedwater system components. This is at best, f
inconvenient, and it could create problems in post-SCRAM response.
- 13. Torus water temperature meter TI 5021-01-A on Panel 903 is difficult to read because of glare and poor contrast.
- 14. PR 3392 on Panel C2 provides information that is inconsistent with operator thinking / expectations.
This recorder displays condenser pressure.
The parameter of interest to the operators is condenser vacuum.
The procedures refer to condenser vacuum. The recorder is labeled condenser vacuum, but it displays absolute pressure (in. of mercury). Operators stated that they would like very much to have this instrument changed.
- 15. A number of annunciator discrepancies were observed:
o No black board; 36 alarms 9 5% power, 219 25% power o
Steam line break in alarm below 30% power
,o Different print size o
Grouping discrepancies and location discrepancies in relation to panel components Only scram and ATWS are color coded o
A number of legends are wordy; some are unclear (e.g., ADS) o o
Small; difficult to read Spelling errors / inconsistent abbreviations o
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o 903L B4 and 903R B1 are identical alarms, core spray break for A &
B.
Nomenclature is different.
- 16. Reliability of indication is a potential problem.
Except for the alarm tiles and the isolation mimic, all control room indication is single bulb, single filament.
The licensee should ensure that redundant control illumination is available, particularly for ES systems which are normally dark.
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4 ENCLOSURE 4 Safety Parameter Display System In-Progress Audit Report for Pilgrim Nuclear Power Station 9
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