ML20202J564
| ML20202J564 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 04/07/1986 |
| From: | Liner R SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| Shared Package | |
| ML20199B041 | List: |
| References | |
| CON-NRC-03-82-096, CON-NRC-3-82-96 NUDOCS 8604160119 | |
| Download: ML20202J564 (169) | |
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i IN-PROGRESS AUDIT OF THE ; DETAILED CONTROL ROOM DESIGN REVIEW FOR CONSUMERS POWER COMPANY'S i BIG ROCK POINT NUCLEAR POWER PLANT N ] April 7, 1986 l t Prepared by: Science Applications International Corporation McLean, Virginia 22102 Under Contract to: l The United States Nuclear Regulatory Commission f
, [ Contract NRC 03 82 096 Post OMce Bos 13D,1710 Goodrktge Orw. McLean, Wginia 221W. (7tt1I C14XO
Ia o Table of Contents (Continued) Section ATTACHMENT D: DCRDR Documentation Controls
;' ATTACHMENT E: Task Analysis - Task Identification (Procedures and Sample Entries)
ATTACHMENT F: Validation - Walkthrough (Procedures and Sample
, Documentation)
ATTACHMENT G: Control Room Inventory (Procedures and Sample Results) ATTACHMENT H: Task Analysis - Instrument and Control Verification
, Worksheets ;
I ATTACHMENT I: Sample NUREG-0700 Annunciator Survey Results L. ATTACHMENT J: Human Engineerina Discrepancy Documentation Procedures 1- and Forms ATTACHMENT K: HED Categorization Procedures ATTACHMENT L: Control Room Design Conventions'for Controls
, ATTACHMENT M: Control Room Design Modification Procedures and Forms ATTACHMENT N: List of Meeting Attendees 6
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v . IN-PROGRESS AUDIT OF THE DETAILED CONTROL ROOM DESIGN REVIEW FOR CONSUMERS POWER COMPANY'S BIG ROCK POINT NUCLEAR POWER PLANT BACKGROUND Consumers Power Company submitted a Detailed Control Room Design Review (DCRDR) Program Plan for Big Rock Point Nuclear Power Plant on June 20, 1984
; (Reference 1). Nuclear Regulatory Commission (NRC) comments on that Program Plan were forwarded to Consumers Power Company on November 7, 1984 (Refer-ence 2). , Based on review of the Program Plan, the NRC staff recommended that a
[ meeting be conducted at the plant to resolve the staff's concerns. Subsequently, the NRC staff conducted an additional review of the subject of the Program Plan. The p rpose of the additional review was to assess the-scopo of the function and task analysis within the DCRDR for determining
, whether a safety parameter display system (SPDS) is necessary for Big Rock l Point. The additional review concluded that the scope of the DCRDR for evaluating the need for an SPDS may be inadequate. The results of this review are presented in Additional Staff Comments on Detailed Control Room Design Review (DCRDR) Program Plan, dated January 22, 1985 (Reference 3).
i In an effort to resolve the NRC staff's concerns regarding the OCRDR ! program at Big Rock Point, a phone conference was held on March 20, 1985. ! The NRC staff and its technical assistance contractor, Science Applications l International Corporation (SAIC), and Comex Corporation (subcontractor to ! SAIC), discussed the DCRDR program concerns with the licensee, and concluded L that the NRC concerns had not been adequately addressed. The minutes of this phone conference are presented in Reference 4. An in-progress audit of the Big Rock Point DCRDR was planned. The purpose of the audit was to clarify certain aspects of the review process, l to confirm that the review was being conducted appropriately, and to i 1 l 9 s
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evaluate the licensee's results to date. ihe,NRC audit plan was forwarded )
; -to the licensee on May 31, 1985 (Reference 5) b l 1 _ ,
' The in-progress audit was arranged through the NRC Project Manager for l Big ' Rock Point and was scheduled for January 27 through January 31, 1986.
The purpose of the audit was.to, compare the organization, process, and res'ul ts of the Big Rock Point DCRDR %ith the DCRDR requirements of Supple-
- ment I to NUREG-0737 (Reference 6)'c
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The 'NRC audit ' team wasfcomprised of the NRC team leader, the NRC project manager, a consuJtin't from SAIC, and a consultant from Comex i f Corporation. The disciplines of human factors engineering and systems
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engineering were represented on the team. The in'-progress audit included review of DCRDR documentation, visits to the control room and remote shut-down facility, and discussions of Consumers Power Company's plans to complete the DCRDR. Attachme$t A provides the audit agenda. Attachment N provides a list of meeting, attendees. ,
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{r In response to the NRC concerns identified during the in-progress. audit, the licensee modified its DCRDR schedule. The licensee's revised
. . schedule was forwarded to the NRC by letter dated March 19, 1986 (Reference 7). NRC/SAIC review of the March 19, 1986, Additional Integrated Plan Schedular Informatio'n, resulted in the identification of several concerns.
Specifically, t'he NRC concerns involved questions about the length of the 3 schedule extension from April 1986 to July 1987, and the Safety Parameter
- Display System Safety Analysis Report submittal.
c These concerns wer'e addressed with the licensee in a March 25, 1986, phone conference. The phone conference was attended by the licensee, the {' NRC, and SAIC. During this conference, the NRC agreed to accept the licensee's integrated liOIng,scheduleprocessasanappropriate mechanism for implementing proposed HED corrections. The licensee committed to submit a Summary Report by March 1;+ 1987, rather than by July 1987. The licensee i[ -also agreed to provide a documented SPDS safety analysis report by March 1, 1987. This repo'rt documents the findings of the in-progress audit and the March 25, 1986, conference call regarding the DCRDR. This report was
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compiled and integrated by SAIC with input from the NRC and Comex Corpora-tion. The report represents the consolidated observations, conclusions, and recommendations of the audit team. DISCUSSION NUREG-0700 (Reference 8) describes four phases of the DCRDR and provides applicants and licensees with guidelines for their conduct. The phases are:
- 1. Planning I
- 2. Review
- 3. Assessment and implementation
- 4. Reporting.
The requirements of Supplement I to NUREG-0737 indicate the need to include a number of elements in the DCRDR. They are:
, 1. Establishment of a qualified multidisciplinary review team.
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- 2. Function and task analysis to identify control room operator tasks and information and control requirements during emergency operations.
i' L 3. A comparison of display and control requirements with a control room inventory.
- 4. A control room survey to identify deviations from accepted human I factors principles.
. 5. Assessment of human engineering discrepancies (HEDs) to determine Ec which are significant and should be corrected.
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- 6. Selection of design improvements.
- 7. Verification that selected design improvements will provide the necessary correction.
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- 8. Verification that improvements will not introduce new HEDs.
- 9. Coordination of control room improvements with changes from other programs such as the safety parameter display system (SPDS),
operator training, Reg. Guide 1.97 instrumentation, and upgraded emergency operating procedures (EOPs). Licensees are expected to complete Element I during the DCRDR's planning phase, Elements 2 through 4 during the DCRDR's review phase, and Elements 5 through 8 during the DCRDR's assessment and implementation phase. '.. Completion of Element 9 is expected to cut across the planning, review, and assessment and implementation phases. A Sumary Report is to be submitted at the end of the DCRDR. As a minimum it shall: ii [ 1. Outline proposed control room changes.
- 2. Outline proposed schedules for implementation.
- 3. Provide sumary justificatis)n for HEDs with safety significance to be left uncorrected or partially corrected:
The NRC staff evaluates the organization, process, and results of the DCRDR. Results of the evaluation are documented in a Safety Evaluation Report (SER) published after receipt of the Sumary Report. An overview of ,;, the Big Rock Point DCRDR activities including schedules is 'provided in l Attachments B and C. See Attachment D for a description of DCRDR documentation. The Sumary Report for the Big Rock Point DCRDR is scheduled to be submitted for NRC staff review by March 1, 1987.
. The audit team concentrated its review on the organization, process, and results of the DCRDR. In-progress audit findings,'in terms of the DCRDR o -elements identified in Supplement I to NUREG-0737, are provided below.
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- 1. Establishment of a Oualified Multidisciolinary Review Team The organization for conduct of 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. The DCRDR team should be given sufficient authority to carry out its mission. A core group of specialists in the fields of human factors engineering, plant operations, instrument and control engineering, and nuclear engineering are expected to participate with assistance as required from other disciplines. Staffing for each technical task should bring appropriate expertise to bear. Human
, factors expertise should be included in the staffing for most, if not all, technical tasks. Finally, the DCRDR team should receive an orientation which contributes to the success of the DCRDR. Section 18.1, Appendix A, of NUREG-0800 describes criteria for the multidisciplinary review team in more detail (Reference 9).
I Review Team Selection. The Big Rock Point Plant (BRP) project organization for the DCRDR was found to be two-layered. The working-level group, referred to as the review team and headed by a manager from the k engineering field office of Consumers Power Company, consisted of six members, augmented as necessary, working on the project full- or nearly-full-time. The output of the review team, assessed HEDs with proposed corrective action, passed to the technical review group (TRG) for cost benefit study and final decision as to acceptance or rejection of HED propo-sals. The authority of the review team appeared adequate to perform an effective review. Orientation and training of the review team and the TRG
. was described as based on previous experience in other job assignments, and extra training in the form of courses, seminars, and similar instruction specifically in the area of human factors for the DCRDR. Some members of the DCRDR review team were also members of the TRG to assist in providing continuity.
Examination of the resumes of the DCRDR team members and discussion concerning member qualifications indicated that each of the disciplines described in Section 18.1, Appendix A, NUREG-0800 was available. The list
- of DCRDR team members is provided in Attachment B of this report. Attach-ment C provides a list of DCRDR activities including the team members 5
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l l l assigned to each activity, and a list of actual man-hours expended by each member for each activity. R.W. Hicks, the DCRDR team leader, is a mechanical engineer with more than fifteen years' experience in fossil plant control room design. He received DCRDR-specific human factors training from Biotechnology Inc. Biotechnology Inc. provided technical assistance to the NRC in the develop-ment of NUREG-0700, Guidelines for Control Room Design Review. D.W. Thayer is a full-time instrumentation and control engineer who also received DCRDR human factors training from Biotechnology. W.J. Trubilowicz received DCRDR
, human factors training from Ger.eral Physics Corp. R.R. Biggs provided j nuclear engineering and operations expertise. P.M. Donnely provided proba-bilistic risk assessment expertise during the DCRDR activities.
t Based on the information presented to the audit team, selection of review team members appears to cover adequately the disciplines, including human factors, necessary to produce an effective DCRDR. Resumes for team members assuming responsibility 'fter submission of the the Program Plan [ and detailed information on course. of instruction and training received by members in the area of human factors engineering will be included in the Summary Report. Manaaement Resoonsibility. The DCRDR review team is managed by a Consumers Power Company employee. The management responsibility of the DCRDR by Consumers Power Company staff is consistent with NRC guidelines and I; policy. l 2. Function and Task Analysis to Identify Control Room Ooerator Tasks and Information and Control Reauirements Durina Emeraency Operations The purpose of the system function and task analysis (SFTA) is to identify the control room operators' tasks during emergency operations and to determine the information and control capabilities the operators need to
~ perform those tasks. An acceptable process for conducting the function and task analysis is:
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- 1. Analyze the functions performed by systems in responding to
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transients and accidents in order to identify and describe those tasks operators are expected to perform. 4
- 2. For each task identified in Item I 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 task performance. Displays and controls are not identified at I this stage. Operator tasks have been identified and described,.
,, and the information and control capabilities necessary to perform 1 those tasks are being determined.
- 3. Analyze the information and control capabilities determined in Item 2 above to determine those characteristics essential to adequate task performance. Information characteristics include parameter type; dynamic range; setpoints; resolution / accuracy;
. speed of response; units; and the need for trending, alarming, etc. Control characteristics include type (discrete or continuous), rate, gain, response requirements, locking functions, I and feedback information associated with control use. ! The described process is nrescriotive. It should identify in detail what operators need to do in order to control the systems which mitigate the consequences of transients and accidents. In addition to its use in the [r DCRDR, it should be the basis for complete and technically adequate I emergency operating procedures (EOPs). Identification of tasks and necessary information and control capabilities should be based on engineering analyses and be independent of displays and controls currently in the control room. Identification of Event Seouences. Function Identification. Function l The audit team reviewed the methodology Analysis. and Task Analysis. employed by BRP to establish operator information and control requirements during emergency operations and noted that it was the same as that described in the Program Plan, Section 4.3. Simply restated here, the process ! consisted of establishing from the probabilistic risk assessment (PRA) (completed and submitted March 1981) 46 accident sequences; examples include 1 7 I
1 l turbine trip, ATWS, small, medium and large LOCA. From these, a set of dominant accident sequences (scenarios) consisting of two or more sequence headings was compiled to ensure that all emergency operations of BRP systems were exercised. The emergency scenarios include the following:
- 1. LOCA and small steam line break
- 2. ATWS
- 3. Spurious reactor depressurization system operation
- 4. Loss of main condenser
- 5. Turbine trip (reactor scram)
- 6. Loss of air
- 7. Station blackout
- 8. Loss of DC power
- 9. Spurious opening of turbine bypass valve
- 10. High energy line break disabling normal core spray Although not required by Supplement I to NUREG-0737, it was noted that startup and shutdown scenarios were also used for a task analysis.
I - Operator action event trees (OAET), as described in the Program Plan Section 4.3.3, were noted by the audit team to have been prepared for each sequence heading (see Attachment E). From the OAET flow chart, a task analysis-task identification worksheet (form BRP-CRDR-11A) was prepared by l corporate DCRDR engineering staff independent of instruments and controls already in the control room. For each operator task on the OAET, BRP-CRDR-g IIA was filled out with information needs, control needs, type of indication or control, scale resolution, etc. (see Program Plan, Section 4.3.4). After the analysis effort to determine requirements, lists were assembled from plant drawings, system descriptions, etc., to determine the expected instru-ment that would fulfill the instrument / control requirement for further entry into BRP-CRDR-11A. BRP-CRDR-11A forms, filled out for each sequence heading, and then gathered for the accident sequence, were packaged into scenario books to be used in the validation effort. In addition, the identified task analysis instrument and control requirements were used in the comparison activity comprising verification as described below. 8
" Validation Walk-through" was the BRP process for determin.ing whether the control room operating crew can perform their tasks effectively in the dynamic sense, given the existing control room instrumentation and controls.
The scenario books of BRP-CRDR-11A forms were used by a team of two to four observers during control room walk-throughs by a normal crew complement. Operators described what instrument or control thgy would use in performing the task listed. If operator choice was different from the expected instru-ment listed / derived from task analysis, the actual instrument used was p listed with coments concerning the alternate choice. Thus, validation produces instrumen? and control requirements thought of by operators, but
,. not necessarily thought of by the independent person performing task analysis. Analysis of all such observations will be performed to determine whether HEDs existed.
The audit team concluded that the above methodology and procedures relative to task analysis completed to date adequately fulfill requirements. This conclusion, however, is based on the licensee's commitment concerning ! completion of task analysis for upgraded E0Ps. Specifically, operator tasks I identified in upgraded E0Ps that were not covered by the selection of
- - sequence headings from the PRA will be similarly analyzed as described p
-above. Further, the audit team understood that task analysis remains to be L completed for the ATWS accident scenario and for the safe shutdown (remote) j panel.
- 3. Comoarison of Disolav and Control Reouirements With a Control Room Inventorv The purpose of comparing display and control requirements with a control room inventory is to determine the availability and suitability of displays and controls required for performance of the E0Ps. Success of this
- element depends heavily on the quality of the function and task analysis and L the control room inventory.
- Display and control requirements should be derived from analyses which are sufficiently detailed to support development
- of complete and technically adequate E0Ps. Characteristics appropriate to the task should be described for each display and control need identified by the function and task analysis. 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 9
0 displays and controls in order to allow meaningful comparison with the results of the function and task analysis. Unavailable or unsuitable displays and controls should be documented as HEDs. The audit team noted that BRP had no preexisting control room inventory of instruments and controls. To establish such a document, a corporate senior designer (I&C) was assigned to use multiple plant sources, including instrument indices, piping and instrumentation diagrams, photos, etc., to prepare a complete inventory (see Attachment G). The inventory format lists the following: F
- 1. Instrument number
- 2. System
- 3. Service (e.g., FD WTR HP HTR Outlet)
- 4. Panel location
- 5. Range, setpoints
- 6. Interconnections
- 7. Type
, 8. Scale resolution I 9. Used (yes, no)
I Utilizing this inventory document, and the output of the task analysis / validation activities, the licensee intends to conduct a comparison ,j activity (task analysis requirements - availability) including a suitability I- determination. The audit team reviewed the " Task Analysis - Instrument and lp Control Verification Worksheet" and noted the following:
- 1. The left section lists task analysis requirements, sourced from the customized BRP-CRDR-llA sheets described above.
- 2. The center section lists " Availability," i.e., instruments available, sourced from the inventory.
- 3. The right section lists " Suitability," where comparison comments are entered, along with the HED number if appropriate.
It was noted by the audit team that this activity will result in comparison of actual control room requirements and controls (inventory) with 10
ea the expected instrument list derived from task analysis and _ validation activities (see Attachments F and H). Noncomparable items result in HEDs (see Attachment J). Based on the description of this methodology given by the licensee, the audit team concluded that regulatory objectives should be fulfilled, and
- that the method is adequate.
- 4. A Control Room Survey to Identify Deviations From Accented Human Factors Princioles r
1 The key to a successful control room survey is a systematic comparison of the control room against accepted human engineering guidelines. One accepted set of human engineering guidelines is provided by Section 6 of NUREG-0700. Discrepancies between the control room and human engineering guidelines should be documented as HEDs. At the time of the in-progress audit, the " static" survey portion of [ the DCRDR was completed. As described in the Program Plan, the survey was I conducted in accordance with NUTAC guidelines, notwithstanding concerns
, expressed by NRC staff in their review comments of the Program Plan.
The audit team evaluated the use of the NUTAC survey guidelines by conducting a survey of the BRP control room in company with plant management i personnel in the area of annunciator warning systems, using NUREG-0700 i , guidelines, Section 6.3. Approximately 25 categories of HEDs were noted by the audit team. The licensee was then requested to summarize all HEDs noted by BRP personnel conducting the NUTAC annunciator survey, or annunciator l HEDs written from other sources; for example, task analysis, operator-questionnaires, or operator interviews. Approximately 10 categories of
; annunciator HEDs were noted by the licensee.
Based on the annunciator survey evaluation, the NRC audit team con-l" tluded that the control room survey is incomplete, and does not meet the Supplement I to NUREG-0737 criteria for surveys (see Attachment I). There-fore, the licensee should perform additional survey activity to ensure that the methodology will achieve results equivalent to the use of NUREG-0700 11
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, guidelines. The licensee is currently developing a methodology for conducting additional survey activity. !
- 5. Assessment of HEDs to Determine Which Are Sionificant and Should Be Corrected HEDs should be assessed for significance. The potential for operator error and the consequence of that error in terms of plant safety should be i systematically . considered in that assessment. Both th9 individual and I
aggregate effects of HEDs should be considered. The result of the
. assessment process is a determination of which HEDs should be corrected because of their potential impact on plant safety. Decisions about whether HEDs are significant in terms of potential impact on plant safety should not be compromised by consideration of such issues as the means and potential I
cost of correcting those HEDs. BRP provided the audit team with forms BRP-CRDR-7 and 7A, entitled Human Engineering Discrepancies, and HED Categorization, respectively (see Attachment K). Form 7 describes the methodology for data entry on the HED form; 7A described in detail the categorization process alluded to in Section 5.2 of the Program Plan: Category I HED is Documented (e.g., LER, questionnaires) Level A - Safety important, did result in unsafe operation or T/S violation Level B - Safety important, did not result in unsafe operation, but could have Category 2 HED Undocumented (e.g., Survey) ! Level A - Safety important, did not result in unsafe i operation, but could have Category 3 HED Documented or Undocumented, not safety related Level A - Not safety.related, but documented Level B - Not safety related, not documented j Level C - Not safety related, did not or could not result in operational error i 12
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- 7. Verification That Selected Imorovements Will Provide the Necessary Correction and Verification That Imorovements Will Not Introduce New ED_s A key criterion of DCRDR success is a consistent, coherent, and effective interface between the operator and the control room. One appro-priate way to satisfy that criterion is through iteration of the processes of selection of design improvements, verification that selected design improvements will provide the necessary correction, and verification that improvements will not introduce new HEDs. Techniques for the verification
, process might include partial resurveys on mocked-up panels, applied experi-ments, engineering analyses, environmental surveys, and operator interviews.
Each iteration of the selection and verification processes should reduce inconsistencies in the operator-control room interface while increasing the coherence and effectiveness of that interface. The consistency, coherence, and effectiveness of the entire operator-control room interface is important
- t. to operator performance. Thus, evaluation of both the changed and unchanged portions of the control room is necessary during the verification processes.
BRP had not established a written, formal methodology for accomplishing r the intent of this Supplement I to NUREG-0737 requirement at the time of the audit. However, the audit team was briefed by the licensee on the iterative process by which final modification drawings were prepared (see Attachment !' L). The process included the preparation of draft drawings from applicable criteria and control room conventions. Consistency is ensured by using i - control room design conventions and specifications such as those illustrated
, in Attachment L and Control Panel Demarcation and Nameplates Guidelines.
The Control Panel Demarcation and Nameplates Guidelines guide the designer in labeling hierarchy, abbreviation conventions, color coding conventions, !' demarcation conventions, letter size conventions, material s, and non-standard label application. , All HED modifications are reviewed by engineering department and opera- ! -tions department personnel prior to implementation. Following implementa-tion in the control room, the final modifications are reevaluated by the design review team in order to verify that the modification has corrected
- l. the HED and not introduced new HEDs. Each modification requires three l-1 15 l
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l ! l Eauipment and Worksoace. BRP is using the control room, the shutdown panel, and photographs of the control room to conduct the DCRDR. No mock-up has been constructed, or is deemed necessary. The DCRDR review team appears to have been provided with adequate references, equipment, and resources to accomplish a satisfactory DCRDR. Schedule. During the March 25, 1986, DCRDR phone conference, attended by NRC, SAIC, and the licensee, the licensee made a verbal comitment to submit the DCRDR Sumary Report to NRC by March 1, 1987. During the phone conference, the NRC also agreed that the licensee's Integrated Living 7' Schedule is an appropriate mechanism for scheduling HED corrections. This
,j- will obviate the need for fixed closure dates on the proposed HED correc-
- tions. However, it is understood that the HEDs will be corrected within the next three refueling outages.
, The licensee also made a verbal comitment to provide an SPDS Safety ,h Analysis Report by March 1, 1987.
['~ All verbal comitments made by the licensee during the March 25, 1986, DCRDR phone conference will be formally documented and submitted to the NRC.
;- This will include a documented comitment to revised DCRDR and SPDS ! schedules.
I CONCLUSION
- An in-progress audit of the Big Rock Point nuclear power plant DCRDR lL was conducted from January 27 through January 31, 1986. During the audit, the NRC audit team compared the organization, process, and results of the Big Rock Po. int DCRDR with the DCRDR requirements of Supplement I to NUREG-0737. The audit included review of DCRDR documentation, visits to the
; control room and remote shutdown facility, and discussions of Consumers L Power Company's plans to compldte the DCRDR. The audit findings are listed j below in order of the nine requirements of Supplement 1 to NUREG-0737.
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- 1. The DCRDR team represents a qualified multidisciplinary review
,.- team which includes adequate human factors expertise. The follow-
.ing documentation should be included in the Sumary Report:
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o . o Four resumes of DCRDR team members. o List of human factors courses attended by the DCRDR team members.
- 2. The system function and task analysis process and results should meet the requirements. Since some of this activity remained to be
; completed at the time of the audit, the following documentation should be included in the Summary Report:
o Verification that the ATWS accident scenario was analyzed. o Verification that the safe shutdown (remote) panel tasks were analyzed. l o Verification that any new operator tasks identified during ~ the development of upgraded E0Ps were analyzed.
- 3. The comparison of the display and control requirements to the control room inventory should meet the requirements.
- 4. The control room survey was found to be incomplete, and did not meet the Supplement I to NUREG-0737 criteria for surveys. The licensee should perform additional survey activity as deemed appropriate to ensure that the methodology achieves results equivalent to the use of NUREG-0700 guidelines. Documentation of
'I the activity should be included in the Summary Report.
- l. 5. The HED assessment methodology and results should satisfy the l [; assessment requirement. Since there are differences between the current assessment methodology and that submitted in the Program
'[ Plan, a description of the current methodology should be included in the Summary Report.
,. 6. The methodology for the selection of design improvements should satisfy the selection of design improvements requirement. The Summary Report should contain:
o Description of Integrated Living Schedule process for HED corrections. i o Description of the methodology for future HED corrections. 20
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- 7. The methodology for verifying that the control rpom changes correct the HED is adequate.
- 8. The methodology for verifying that the control room changes do not introduce new HEDs is adequate. The documented methodology for
, verifying that the control room changes correct the HEDs and do l l not introduce new HEDs should be included in the Summary Peport.
- 9. The coordination of the DCRDR with other control room improvement programs such as SPDS, Reg. Guide 1.97, operator training, and
. upgraded E0Ps should meet this requirement. A description of the coordination activities should be provided in the Summary Report.
In summary, the licensee is in the process of completing a DCRDR which should satisfy the requirements of Supplement I to NUREG-0737. The DCRDR -
- r. Summary Report is scheduled for submittal to NRC by March 1, 1987. The licensee's Integrated Living Schedule will be used to satisfy the Supplement I to NUREG-0737 requirement for a control room modification schedules. The SPDS Safety Analysis Report will be submitted to NRC by March 1,1987.
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REFERENCES
- 1. Letter to Dennis M. Crutchfield, NRC, from: R.M. Krich, subject: Big
. Rock Point Plant - Control Room Design Review (CRDR) Program Plan, June 20, 1984.
- 2. Letter to: David J. VandeWalle, Consumers Power Company, from: John A. Zwolinski, NRC, subject: Staff Comments on Big Rock Point Detailed Control Room Design Review (DCRDR) Program Plan, November 7,1984.
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- 3. Letter to: David J. VandeWalle, Consumers Power Company, from: John
! A. Zwolinski, NRC, subject: Additional Staff Comments on Detailed Control Room Design Review (DCRDR) Program Plan, January 22, 1985. ; 4. Letter to: Leo Beltracchi, NRC, from: Joseph DeBor, Science Applica-tions International Corporation, subject: Detailed Control Room Design Review Evaluations, Phases III-V, Big Rock Point Meeting Minutes for March 20, 1985, Detailed Control Room Design Review Phone Conference, March 25, 1985.
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- 5. Letter to: John Zwolinski, NRC, from: William H. Regan, NRC, subject:
I. In-Progress Audit Plan for the Big Rock Point Detailed Control Room Design Review, May 31, 1985.
- b. 6. Supplement I to NUREG-0737 - Requirements for Emergency Response Capa-bility (Generic Letter No. 82-33), U.S. Nuclear Regulatory Commission, December 17, 1982.
[ 7. Letter to: Director, Nuclear Reactor Regulation, NRC, from: Ralph R. Frisch, Consumers Power Company,
Subject:
Docket 50-155 - Big Rock . Point Plant - Additional Integrated Plan Schedular Information, March 19, 1986.
- 8. NUREG-0700, Guidelines for Control Room Design Reviews, U.S. Nuclear Regulatory Commission, September,1981.
- 9. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 18.1, 22
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,e ATTACHMENT A In-Progress Audit Agenda 1
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i p E [. g'g ~ c UNITED STATES RECEIVED
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; .1 NUCLEAR REGULATORY COMMISSION orow.o.c. . . JVN 5 1985 1
%,*****/ ItAY a 1585 sm ucensing Project.0ffice MEMORANDUM FOR: John Zwolinski, Chief Operating Reactors Branch No. 5 Division of Licensing '
FROM: William H. Regan, Jr., Acting Chief j Human Factors Engineering Branch Division of Human Factors Safety i
SUBJECT:
j IN-PROGRESS AUDIT PLAN FOR THE BIG ROCK POINT DETAILED CONTROL ROOM DESIGN REVIEW The staff plans an In-Progress Audit of the Big Rock Point Detailed Control
- Room Design Review. The purpose of the audit is to clarify certain aspects of the review process, to confim that the review is being conducted appropriately, and to evaluate the licensee's results of the review to date. 'l Our audit plan is enclosed and we request that it be forwarded to the licensee.
4 Based on infomation from the licensee, we suggest the audit be conducted during the dates of September 3-6, 1985. Please confim this schedule by August 1,1985 in order that we may complete travel plans and infom our
; contractor.
The scope of our audit includes the Detailed Control Room Design Review and the Safety Parameter Display System. The staff, with the assistance of a contractor, will conduct the audit in the sequence described in the enclosed plan. Because of the scope of the audit, we anticipate it will require a full four days of effort. In preparation for the audit, please request that the licensee make the following arrangements to facilitate the review process:
- 1. Make available any of the licensee's review team personnel knowledgeable of the DCRDR tasks / processes and be ready to discuss or respond to NRC Audit Team questions during the audit.
! 2. Be able to provide documentation of the review results to date, r
including HEDs identified, and all results from review-products, such as task analyses, surveys and/or inventory checklists, data collection foms, etc.
Contact:
L. Beltracchi x24879
i' . . r-John Zwolinski NAY 31 1585
- 3. Provide for access to the Big Rock Point control room area for the NRC staff and its contractors. The NRC staff also requests clearance to bring camera equipment on-site to take photographs in the control room and other areas which may assist in evaluating the DCRDR and help in documenting the audit.
2
- 4. Provide for conference or office space for both the NRC Audit Team's meeting with the licensee's Review Team and for the NRC Audit Team to meet privately. ,
nl ,J William H. Regan r ef Human Factors Enginep/ /ing B anch Actt { Division of Human Fay, ors Safety i
Enclosure:
j As stated cc: C. Miller . R. Dudley R. Liner, SAIC ' R. Peterson, LLNL [- {. ' l I
-h .
. _ . . . - . - - - , - - - - . ,. . . - . = . - . - , - - . . - . . - . - - - - - - - - - - - - - -- --
AUDIT PLAN DETAILED CONTROL ROOM DESIGN REVIEW IN-PROGRESS AUDIT BIG ROCK POINT
, A. ENTRY BRIEFING:
- 1. NRC
- a. Provide background for audit:
i State NRC requirements d Restate staff :omment on Program Plan
- b. Define and discuss agenda
- c. Discuss reporting of audit results.
~
- 2. Big Rock Point
- a. Provide status of DCRDR Define modifications to Program Plan, if any
~
b.
- c. Define and discuss audit schedule. .
B. CONTROL' ROOM ORIENTATION (Conducted by licensee)
-l if 1. Brief tour of control room for NRC Audit Team
- 2. Definition and location of engineered safeguard systems, including controls and instrumentation -
- 3. Definition and location of instruments to monitor the following critical safety functions:
- a. Reactivity Control, I Reactor core cooling and heat removal to the isolation o b.
condensers,
,' c. Reactor core coolant system integrity,
- d. Radioactivity control, i- e. Containment conditions.
l l
i l 1
- ,- 2
- 4. Definition and location of instruments used to monitor entry symptoms to emergency operation procedures C. AUDIT ELEMENTS OF DCRDR
- 1. Evaluate the DCRDR Team qualifications and the DCRDR Program
- a. Qualifications of members of the review team
- b. Need for formal program of HF orientation and training for non-HF specialists I c. The role, responsibilities, and level of effort of the humar.
factors specialists, and for other team members
- d. Documentation and document control
- 2. Audit methods used in the Function and Task Analyses:
! a. Detailed description and discussion of how the operator's information and control needs and their associated characteristics are determined
- b. Evaluate records and documents associated with (a) e 3. Control Room Inventory:
Evaluate methodology for comparing existing control room instrumentation with required display and control characteristics i
- 4. Audit Control Room Survey:
[. a. Evaluate survey lists and checklists
- b. Evaluate survey procedures
- c. Evaluate survey team
- d. Evaluate review of Remote Shutdown Panel
- 5. Audit HED Assessment:
- a. Review methodology, criteria, and guidelines used in assessment procesi.
v b. Evaluate results from assessment process
~~
- c. Evaluate schedule of HED corrective actions
j . t i
- 6. Audit Selection of Design Improvements:
- a. Review methodology used to identify design improvements
- b. Audit specific design improvements
- c. Evaluate role of human factors specialist in the design-improvements process
,[ 7. Audit Verification Activities that Illustrate HEDs are Corrected: 7
.a. Evaluate methodology used
- b. Evaluate results from verification activities
- 8. Audit Verification Activities Which Illustrate that Improvements Do i Not Introduce New HEDs:
- a. Evaluate methodology used
[
- b. Evaluate results which illustrate that improvements do not introduce new HEDs
- 9. Audit Coordination of the DCRDR With Other Supplement 1 NUREG-0737 Initiatives:
l *- a. Evaluate cocrdination methodology p b. Evaluate results from coordination effort I
- 10. NRC Audit Team Evaluation of One or More Panels in the Big Rock Point Control Room:
- a. Conduct a benchmark evaluation of licensee's review process by independently identifying HEDs i b. Request Big Rock Point to show where HEDs defined in Step (a)
- were found in the DCRDR
- 11. Review of Activities Which Impact DCRDR:
l
,. a. Results from Probabilistic Risk Assessment l '; b.. Results from operator survey (if applicable) l l
l i
I - e
. 12. Review Big Rock Point's Results on the Need for a Safety Parameter Display System:
- a. Evaluate methodology used
- b. Evaluate results in terms of the physical location of instruments within the control room D. EXIT BRIEFING
- 1. NRC Audit Tean:
- a. State preliminary findings from audit
- b. Discuss reporting of Audit Results
- 2. Big Rock Point:
- a. Comments on preliminary findings.
e f 6 F. l l i i- ' l l l f
e . V' 4 i ATTACHMENT B DCRDR Overview Presentation Slides 9 t L t' t.. i . 4 6 I I i . l
-~ ' ~~ ~ ' ~ ' ~ ^- ^
_ . _ . . _ . ~_., _ _ ~ZZ " ' ~ ~ . . _ ., n.r d 4 . j PRO _ JECT ACTIVITIES A. PLANNING 1. SELECT TEAM MEMBERS ! 2. DEVELOP PLAN 1 I B. EXECUTION 1. SURVEYS l
- 2. QUESTIONNAIRES l STATIC j 3. CHECKLISTS
! 4. TASK ANALYSIS 1 l 5. CONTROL ROOM INVENTORY L DYNAMIC ! 6. VALIDATION (
- 7. VERIFICATION '
C. ASSESSMENT 1. GENERATE HED'S l
- 2. CATEGORIZE i
- 3. PROPOSE CORRECTIVE ACTION 1
- D. IMPLEMENT 1. ASSESSMENT AND RANKING BY TRG
- 2. COST BENEIFT
- 3. SCHEDULE BY RANKING i
l
- 4. COMPLETE CORRECTIVE ACTION E. DOCUMENT 1. FINAL REPORT
) B. EXECUTION
- _ STATIC
- HISTORICAL DATA OPERATING EXPERIENCE
! ENGINEERING EXPERIENCE CONTROL ROOM OVERVIEW i LABEL, MIMIC AND DEMARCATION i GENERAL PANEL ABBREVIATION / ACRONYM ' COLOR CODING ! CONTROL DEVICES l REPETITIVE GROUPS ANTROPOMETRIC l COMMUNICATIONS : , ANNUNCIATOR j LIGHTING j SOUND 1 ENVIRONMENT DYNAMIC: TASK ANALYSIS VERIFICATION CONTROL ROOM INVENTORY : VALIDATION l NOTE: TASK ANALYSIS BASED ON PRA QPERATOR ACTION j EVENT IREES (OAET) i l ! t 1 m
.: ___ -_ __ __ _ _T . . _ _ _ _ . _ , . _ _ _ _ .
l PRO _J_ECT STATUS PHASES _ STATUS PLANNING G COMPLETE l EXECUTION S IN PROGRESS ASSESSMENT G IN PROGRESS IMPLEMENT S ENHANCEMENTS IN PROGRESS FINAL REPORTS 9 MODIFICATIONS NOT STARTED O GATHERING DATA
STATUS 1 4 A. PLANNING: CURRENT TEAM l 1 ' R.W. HICKS
- D.W. THAYER*
- FULL TIME R.E. SCHRADER 4
R.R. BIGGS i P.M. DONNELLY t W.J. TRUBILOWICZ
+ 3 MORE AS NEEDED 4
- PLAN
- DEVELOPED AND SUBMITTED JUNE 20, 1984 - PER NUREG 700 l ANY REVISIONS TO PLAN WILL BE INCLUDED IN j
FINAL REPORT ' SECTION 5.2.1.1.3 1
, - - s .. . . _ __
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' 4 , ,
C. ASSESSMENT .
! HED GENERATION - 4 STEPS ,
\
. ~.
- 9 FROM SURVEY, CHECKLIST, VALIDATION, VERIFICATION, ETC. IDEN-l .
A. TIFY DISCREPANCY , ,; N- , , S RECORD DATA AND ADD OPERATIONS INFORMATION S CATEGORIZE BASED ON PREDETERMINED CRITERIA ' ' a
, 9, ~ PROPOSE CORRECTIVE ACTION AND SUBMIT TO TRG .(TECHNICAL j
REVIEW GROUP) FOR RANKING AND INCORPORATION INTO BRP L!VING { SCHEDULE , ( , d a 4
i D. IMPLEMENT l (PLANT FUNCTION) i l 9 ONCE TRG RANKS CORRECTIVE ACTION AND IT IS PLACED ON LIVING SCHEDULE - A MODIFICATION WILL BE MADE S THE MODIFICATION (MOD) WILL BE CARRIED OUT UNDER NORMAL PLANT ! PROCEDURES i 9 ONCE PLANT IDENTIFIES NEED FOR MODIFICATION - CRDR TEAM RELIN-l QUISHES CONTROL TO PLANT AND PLANT WILL CLOSE OUT THE HED 1 l l l
_ _. _~ . . . _ ._. __. __ . _ . . . . ._ J J. t i ! E. DOCUMENT 1 ANTICIPATED l (WILL REQUIRE PROGRAM PLAN REVISION) 6 9
SUMMARY
OF HED'S HER DESCRIPTION CAAGORIZATION PROPOSED CORRECTIVE ACTION l SCHEDULE WHERE AVAILABLE
- 9 PROGRAM PLAN MODIFICATIONS S SPDS JUSTIFICATION
- (IF NECESSARY) ,
i i i
~
l TASK ANALYSIS INPUT l OPERATOR _
~
IDENTIFY INDIVIDUAL ACTION SEQUENCE HEADINGS EVENT TREES (OAET) o FEEDBACK _ FLOW CHART SEQUENCE FROM ~ TO ACCOMPLISH HEADING VALIDATION TASKS u
- DEVELOP WRITTEN i
TASKS AND NEEDS h ORDERS TASKS AND NEEDS FROM FLOW CHARTS OUTPUT u VERIFICATION , ASSIGN ' EXPECTED' INST. TO DO TASKS VALIDATION
b 1 I
- VALIDATION 1
l INPUT l l i TASK ASSEMBLE SEQUENCE ANALYSIS = HEADINGS PER PRA PRA DOMINANT ACCIDENT 7 SEQUENCES (>10 / YEAR) I { CUSTOMlZE SUCCESSES l AND FAILURES n i
; PERFORM WALKTHROUGHS l OUTPUTS EOP , CONSOLIDATE OBSERVER TASK ANALYSIS COMMENTS VERIFICATION ; HED 1
I
1 . l VERIFICATION j INPUT 3 TASK -. r ASSEMBLE LIST OF ANALYSIS EXPECTED INT. l VAllDATION - - - - - - - - - o CONTROL l COMPARE ' EXPECTED' ROOM = TO ' ACTUAL' INST. INVENTORY IN CONTROL ROOM OUTPUT HED
' ~
. f . . . - .- .. . . .- . - .
1 SPDS (ELEMENTS) 737 I NUREG TASK ANALYSIS ENHANCEMENTS 4 0835 s SPDS N OPERATOR - " VERIFICATION EXPERIENCE ANALYSIS ! OPERATOR % VALIDATION l INTERVIEW HEDS i 4A&4C (ALL) 1 OTHER UTILITY INPUTS (REPORTS, SEMINARS, ETC.)
- EXISTING INST. (CRI)
~' ~
-. ._ _ i ~ : 1__ -
. ~_.. _ __ . . . . ~ 1 .._ f 4
PROCESS 737
- VARIABLES F_ UNCTIONS REACTIVITY (POWER) REACTIVITY RX PRESS CORE COOLING AND HEAT REMOVAL l
CONT. TE P RAD OACTIV l PRESS CONTAINMENT CONDITIONS 1 i LEVEL i . ! OFFGAS/ STACK GAS I i 4
f' *
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4 ATTACHMENT C [. - Big Rock Point CROR Schedule Update Activity Status 4 e 4 t: i f I- ). l
f Q PDonnelly, Big Rock RR81ggs, P26-407 T: WJTruollowicz, Big Rock RESenrader, Big Rock Fac, RWHicks, P13-426 [D$/ l May 10, 1985 Dave pgy(g[ SVestet BIG ROCK POINT PLANT CRDR SCHEDULE UPDATE ACTIVIIY STATUS "' c"o n s t e m' .: t . : t CC GHRPetitjean, Big Rock RMKrich, P24-615
, ACSevener, Big Rock NLHaskell, P26-204B Hick-10-85 DPBlanchard, Big Rock JAJohnson, Big Rock JLKuemin, P24-503 RBuckner, Big Rock
- , 72
- ::: :: ' : 2,':::::,': 2 0 0, _
740/42*50*10*ll/L Attached is Revision 4 dated May 10, 1985 of the CRDR Program Schedule and task activity breakdown. The schedule has significantly changed since Rev 3 was issued on October 18, 1984. Specifically, there have been changes in the Company personnel assigned to complete the project, more detail is provided for remaining activities, and project completion has moved into 1986. Significant milestones completed or in progress include:
- 1. Sympton Action Sets completed which provide input to initiate Task Identification and I&C verification activities.
j, 2. Panel drawings are being prepared to include demarcation and labeling changes resulting f rom the Control Room Survey and Design Convention Survey. Short term objectives include: l' l. Develop task identification survey form and conduct survey. I.
- 3. Complete panel drawings and submit for review.
l! I', 3. Initiate Human Engineering Discrepancy (HED) form development and record prior identified HEDs. j 4. Enter HEDs into automated data base program for future cross l referencing and catagorization. l l} It should be noted that even though a project team transition phase has been included on the Program Schedule, none of the activities currently in progress have been af f ected and it appears disruption to the project will be minimal. l i l
- -~
= 198L = 'c85 = .986 V < < A S J \ J < V A V
< < A S E \ J < V A V e
i l 'i l l 1 3 5 ' l s 11 12 15 16 17
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i REFUELING PROJECT TEAM TRANSITION i RP CROR MOGRAM SCHEDULE EV 4 5/10/85 Illd,16ste lt se er s se 6VI
e 1 i ATTACHMENT B BRP CRDR Program Schedule (Rev 4, 5/10/85) Activity Activity Team Activity Number Descriotion Member Status 1 Historical Data Review Completed 2 Operating Experience Questionnaire Completed 3 Follow up Interviews Completed 4A Engineering Dept. Questionnaire Completed 4B Operating Survey Questionnaire completed 5 Survey Checklists Completed 6A Design Convention Surveys Completed [- 6B Measurement Surveys
~
Completed 7 Sympton Action Sets __ Completed 7 8 Task Identification RWH/DWT 6/1/85 Develop form, define accident sequences 2#D WJT from OAET by defining inputs, through- RRB puts and outputs by which actions are ; implemented. (No HED's) _ 9 Verification of IEC RWH/DWT 7/1/85 Develop form, based on activity 8 results- WJT inventory control room to assure existence of PMD(Review) instruments and controls, cross check prior RES HED's for conflicts, record new HED's - f 10 Walk Throughs RWH/DWT 9/1/85 Develop forms (ine summary form) conduct walk- WJT throughs based on those required by NUREG 700 RRB (3.8.2) using scenarios as described in the PDM OAET. Validate task identification analysis (Operators) and I&C verification analysis results. RES
- g. Interview operators regarding layout, arrangement and design, information for HED classification input.
1 Doc ID 8721A
t-4 Activity Activity. Team Activity Number Description Member Status 11 Assessment of HED's RWH/DWT 6/1/85 Compile HED summary, write up prior identi- PMD RRB fled HED's, review and/or modify methodology WJT j .to catagorize HED's, photograph each HED, l develop photograph management system. Categorize HED's (on going). 12 Propose HED Corrective Action .RWH/DWT 11/15/85 Proposed corrective action logged on HED RRB PMD form. Resolution can be in form of paint, WJT DPB
, tape and label (static), backfitting (which may include SPDS), procedures changes and training. Request CRMP perform PRA cost benefit. Analysis on major HED's. Submit summary of proposed resolutions to TRG for , further assessment and ranking within plant priorities.
13 Develop ongoing HED Program RWH/DWT 1/1/86
! Establish procedure to use in future pro- RES l jects incorporate requirements in appro- -
priate plant documents assuring compliance j (FC, SC, SPC, etc). V 14 Develop SPDS Justification RWH/DWT 11/15/85 j Utilize relevant data extracted from appro- DPB (Tentative) priate documents to develop justification RES (Review) I (for or against). Sources: Operator interviews, walkthrough, HED's. Write justification, submit for review to TRG and i {' i incorporate into final report. 15 Compile Documentation & Draft Final Report RWH/DWT 1/1/85 (later) . ., 16 Final Report Review and Comment RWE/DWT 3/15/86 i Submit to NRC RRB PMD ' i (later) WJT RES JLK l* l 17 Project Close Out RWH/DWT 5/1/86 (later) i! i . 2
- Doc :D 8721A
f t Activity Activity Team Activity Number Description Member Status 18 Prepare Panel Drawings DWT/RRB 3/15/86 i CAD panel drawings, include demarcation and labels, add to panel drawing series, reivse nameplate list. Compare with HED's. 19 Review Panel Drawings DWT 6/15/85 Submit to team members and plant for review, WJT RES discuss with operators. 20 Revise Panel Drawings DWT 7/15/85 Based on feedback change drawings RRB RES
, 21 Implement
! RWH/DWT 8/1/85 1
' Arrange for label and material to place WJT/RRB static changes on panels. Determine who, REO 1 ,
how and when. Verify completed per drawings. Implementation will be through use of BRP- FC or SC. - I
. I,4 i
i .. a i 3
, Doc ID 8721A
p g,,. _ . . _ _ , . KAT 01-85 ATTACHMENT B BRP CRDR SCHEDULE MAN-HOUR ESTIMATE (REV 3. 01/07/85)
! Activity Scheduled Team Man I ; # Activity Description Duration Member
- Hours Time Activity Status j 1 Historical Data Review 2 wks KAT 50 63 Completed on FT 32 40 05/11/84 HGB 4 5 WJT 4 5 2 Operating Experience 4 wks KAT 40 25 Completed on
! Questionnaire WJT 10 6 08/20/84 3 Follow-Up Interviews 2 wks KAT 60 75 Completed on RRB 40 50 09/20/84 i WJT 4 5 f 4A Engineering Department 2 wks KAT 8 10 Completed on j Questionnaire (RES) 32 40 06/25/84 4B Operating Survey 3 wks KAT 40 33 Completed on Questionnaire WJT 10 8 10/15/84 j SA Design Convention Surveys 11 wks KAT 200 45 Completed on
. RRB 80 19 08/31/84 l
i 5B Measurement Surveys 2 wks KAT 40 50 Completed on RRB 40 50 11/15/84 (SP&LS) 32 40 6 Survey Checklista 6 wks KAT 140 58 Scheduled for (including human factors RRB 80 33 completion on
, static review of begh giE'"IE WJT 16 7 02/01/85
+pmee monitors panel) FT 16 7 jpovo- ,
IC0185-00048-MS01
. .., . . . . - . - - . - - - -- -~ ~ ~ ~ ~~'
Prge 2 cr o Activity Scheduled 1 Team Man %
# Activity Description Duration Member
- Hours Time Activity Status
) 7 Prepare Preliminary 4 wks KAT 100 63 Scheduled to
; labeling. Demarcation RRB 100 63 commence on i
and Mimic Designs for
- Panels and Console 02/01/85 8 Prepare Summary Listing 2 wks KAT 40 50 Scheduled to of HEDs Documented to Date commence on l
01/15/85 ' l 9 CRDR Team and Plant 4 wks i KAT 8 5 Scheduled to Operations Review and RRE 8 5 commence on Comment on Activity #7 WJT 16 10 Designs 03/01/85 FT 16 10 HCB 8 5
-(Plant ops) 60 38 10 Incorporate Comment into 2 wks KAT 40 1 50 Scheduled to Activity #7 Designs RRB 40 50 commence on i 04/01/85 j II Implement Labeling, 5 wks KAT 8 Mimics and Demarcation 4 Scheduled to
' RRB 8 4 comumence on
} According to Activity WJT 24
#10 Designs 12 04/15/85 (Plant ops) 100 50 12 Sympton Action Sets 10 mos HCB i
Completion by 1 1 13 Task Identification 2 wks KAT 50 63 Scheduled to I WJT 50 63 commence on RRB 50 63 04/15/85 i HCB 50 63 1 i 4 3 i i
- IC0lH5-0004R-NS01
+
To MGBasydlo, Big Rock WJTruh110wicz, Big Rock RRBiggs P-26-407 FATurski, Big koek Prom KAToner P-24-116 CONSUMERS POWER Date January 7, 1985 COMPANY Subject BIG ROCK POINT PLANT Internal CONTROL ROOM DEdIGN REVIEW (CRDR) Correspondence PROGRAM SCHEDULE - ACTIVITY STATUS CC KWBerry, P-24-100A MMVanWagner, Palisades KAT 01-85 PFBruce, Palisades DPBlanchard, Big Rock-GERPetitjean, Big Rock JAJohnson, Big Rock 4 ACSevener, Big Rock WGMerwin, Big Rock
! RMKrich, P-24-615 DCC: 740/00250/50000/L JLKuomin P-24-503 750/42*50*10*11/L Attached is a revised CRDR Program Schedule and Manhour Estimate. The schedule and estimate reveal the following important aspects of this project:
- 1. The project schedule has slipped approximately four months as evidenced by the final report to the NRC scheduled to be submitted by 10/21/65 instead i of 06/17/85 (refer to KAT 49-84 for revision 2 of schedule). The primary reasons for this slippage are:
' -j' 2
- a. A delay in completion of the Symptom Action Sets (activity #12) of approximately 2 months (the sets are required to perform operator walk-throughs of various accident scenarios in an attempt to determine i the capability of existing control room instrumentation and control to support accident mitigation efforts).
- b. The inclusion of additional project activities to re-label, demarcate and re-mimic as necessary the console and panels as soon as possible l (these activities are numbered 7 through 11) as requested by upper
; management.
lf l c. Resource requirements of the EEQ program which bid for attention l of certain CRDR tesa members (the EEQ project takes priority over the j CRDR as directed by TRG ranking).
- 2. The human factors review of the high range samma monitors will be conducted 1 in two phases; a static review of the instrumentation against human
! factors guidelines will be completed by 02/01/85 as part of the panel survey checklists (activity #6), and a dynamic review of the capability of operators to use the instrumentation in accident conditions will be completed by 06/15/85 as part of the operator walk-throughs. These ! reviews are required by the Atomic Safety and Licensing Board and have ! been implemented into the CRDR project appropriately.
- 3. It is requested as part of activity #11 that the Plant Operations Department j install (or at least assist in the installation) the new labels, simic and l demarcation based on the designs provided by the CRDR team. This additional i
support is necessary since the team will be engaged in other CRDR activities j during this period of time. 4 l IC0185-0004A-NS01
I i' ATTACHMENT D i DCRDR Documentation Controls f {' e
i *
' ~
DOCUMENTATION CONTROLS BRP-CRDR-1 Historical data review record BRP-CRDR-2 Operating experience questionnaire BRP-CRDR-2A Questionnaire items summary BRP-CRDR-4 Operator experience follow-up interview BRP-CRDR-4A Follow-up questionnaire BRP-CRDR-4C Follow-up summary l BRP-CRDR-5 Engineering questionnaire BRP-CRDR-6 Control room survey operator questionnaire o BRP-CRDR-8A Overview checklist ;- BRP-CRDR-8B Operator assisted checklist 3 BRP-CRDR-8C Labe'-demarcation-mimic checklist m i T: be completed after 85 outage BRP-CRDR-8G General panel checklist BRP-CRDR-9A Abbreviations / acronym survey BRP-CRDR-9B Color coding BRP-CRDR-9C1 Control Devices 9C2 Meters BRP-CRDR-9D Repetitive Group BRP-CRDR-9E Anthropometric i BRP-CRDR-9F Annunciator
,' BRP-CRDR-9G Communica'tions -
BRP-CRDR-10 Symptom Action Sets ' BRP-CRDR-11 Task Identification BRP-CRDR-12 Task Verification I u BRP-CRDR-13 Control Room Inventory ;; BRP-CRDR-14 Task Validation (walk-through) g
- BRP-CRDR-7 Human Engineering Discrepancy i With corrective action and categorization!jy I
l 1 s 4 i
FORM BRP-CRDR-1 STATUS I have completed a review of BRP-CRDR-1 file (historical data review). The a required form BRP-CRDR-1 has been generated and information required by that form has been filled in. A summary has been performed by virtue of a cover letter (CPCo IOM KAT 27-84 dated May 21,. 1984) which provides a summary list of potential problems (Attachment 1). The summary list is further expanded by Attachment 2 which contains completed BRP-CRDR-1 forms sequenced from one to twenty-four which explain specific details of potential problems. All twenty-four of those items are reflected in BRP-CRDR-4, Interview Questions, and therefore no HEDs need to be generated from this portion of the CRDR. l 4 i l i 1. l i. W A Staley 10/18/85 i Doc ID 9185A t
- - - - - - - - , - - - - . . e.---n--- -- - , , , n-- - - - - , - - - . - . - - - . - . , , - - - . - , , - , - - - - . - - - - - - - - - - - - . , - - - , - - ,,
FORM BRP-CRDR-2 STATUS I have completed my review of BRP-CRDR-2 file (operating experience questionnaire). I f ound that the required f orm BRP-CRDR-2 has been generated, and that the form which includes a biographical and physical data sheet along with an operating experience questionnaire, was distributed to the licensed operators at BRP. Nine operators returned those forms to be included in the CRDR. Due to the voluntary nature of the request for desired information, not all of the forms are completely filled out. I do not believe this is a problem because a follow-up was performed by virtue of BRP-CRDR-4. A summary of the BRP-CRDR-2 forms was completed and is documented on form BRP-CRDR-2A. This summary tabulates all of the responses from the operating experience questionnaire into positive, negative and neutral responses. The positive responses mostly provide yes or no answers, but they do contain some potentially useful information. The neutral responses were essentially unanswered questions. The negative responses however, would normally require that an HED be written against them for consideration at the assessment phase. In this case no HZDs will be written because all of the negative responses were compiled with questions to form a questionnaire for the
- *0perating Experience Questionnaire Follow-Up Operator Interview" form BRP-CRDR-4 A. Any HEDs that would have been written for CRDR-2 will be clarified in CRDR-4 and if necessary generated as part of that section's requirements.
; I also conducted an audit to insure that all operator responses were
- considered on the summary and audited even further to insure that all of the negative responses were carried forward to the follow-up interview form. I found no discrepancies in the transfer of information.
,. A summary of positive responses will be compiled and sent to file BRP-CRDR-2
{ and to the assessment team for their consideration at the assessment phase. I
.o- l
/
W taley 10/16/85 l' i i Doc ID 9185A I
\
i
i o s. I' FORM BRP-CRDR-4 STATUS This form actually consists of four sections. They are:
- 1) BRP-CRDR-4 (Operating Experience Review Follow-up Operator Interview Q uestions), 2)
BRP-CRDR-4A (Operating Experience Questionnaire Follow-up Questions), Operator I nterview Response Summary) and 4) BRP-CRDR-4C, Part B3) BRP-CRDR-4C, Operator Response Summaries). (Follow-up Operator Interview BRP-CRDR-4 B . I believe that this form was inadvertently omitted.There is no during those interviews.For BRP-CRDR-4 eleven operators were e out interviewed and that' form was also filled out during those interviews.For BRP-CRDR-4 By virtue of the
; operator accomplished both interviews. names and dates on the two for l - t BRP-CRDR-4 and BRP-CRDR-4C, Partrgathered during BRP-CRDR-4 A interviews.
Bof is a summary I generated summaries. 97 HED's as a result of keh hviewing BRP-CRDR-4 & j 4A and their will be when I complete the reviews for all of theThestatic CRDR HED's will then be forwarded to R W Hicks. . t i I t 0 j :dWf WA taley 11/14/85 Doc ID $9185A
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- O FORM BRP-CRDR-5 STATUS I have completed my review of BRP-CRDR-5 (Engineering Questionnaire) with the following results: The form has been generated and was filled out by the Engineering Department representative. A summary is not necessary because all necessary information is contained on one form. No HED's were generated so I generated the necessary HED's for this form (36).
I i f I i i I i WAStaley 1 11/26/85 Doc ID $9185A
f FORM BRP-CRDR-6 STATUS I have completed my review of BRP-CRDR-6 (Control Room Survey Operator Questionnaire). The required form has been generated and was distributed to the licensed operators. Ten operators filled out and returned the form to the review team. A summary of this form was not compiled. I compiled a summary using the existing BRP-CRDR-6 f orm and titled page 1 as ' summary". That summary was completed on 11-15-85. No HEDs have been issued against this form as of 11-15-85. Af ter completing a review of the summary, a total of 85 HEDs were generated to close out the review of BRP-CRDR-6. i [ i 4 l l . i t t t I WAStaley 11-19-85
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f FORM BRP-CRDR-8 STATUS This form consists of four parts, they are: 8A Control Room Overview Checklist, 8B Operator Assisted Checklist, 8C Label, Mimic and Demarcation Checklist and 8D General Panel Survey Checklist. All four of these forms have been generated. Forms 8A, 88 and 8D have been filled out. Form 8C will be filled out and HED's generated after the control room demarcation has been completed. A summary is not deemed necessary for these checklists. The data for these forms has been compiled so that discrepancies are readily identifiable and HED's can be easily written without a summary. HEDs were generated as the 8D checklist was filled out by members of the review team for that portion. I generated the HED's required by forms BRP-CRDR-8A and 8B. i i$ e - i i I
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l [ i WAStaley/ 11/20/85 Doc ID 49185A i
FORM BRP-CRDR-9 STATUS This form is comprised of eight sections. They are Acronym survey), 98 9A (Abbreviations and (Color Coding Survey), 9C1 (Control Devices survey), 9C2 (Meters Survey), 9D (Repetitive Groupings Survey), j 9F ( Annunciator Survey) and 9G (Communication Survey).9E All of(Anthropometric the above forms Survey), l I have been generated and the surveys completed. Forms 9A, 98, 9C1, 9C2, 9D and 9G have summaries for their repsective files. Forms 9E and 9F do not have summaries but data sheets were compiled as necessary to summarize their information. Forms 9A and 98 are no longer current and another survey will be completed to reduce the number of HED's necessary for these forms after demarcation and painting is complete. A review of 9C1 generated 32 new HED's. Forms 9C2 and 9D are complete and require no further action. required 21 HED's to be generated. Form 9F required 12 NED's to be Form 9E generated. Form 9G is out of date due to changes made in communic ~ation equipment and another communication survey should be completed. f I i! 4 6 l j l ,l li t i! r l l WAstaley A 12/5/85 + Doc ID 09185A 1
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I ATTACHMENT E Task Analysis - Task Identification (Procedures and Sample Entries) l 1. I l. i + 7
.r BRP-CRDR-ll Page 1 of 2 TASK ANALYSIS - Tk's K IDENTIFICATION Task identification involves two steps. The first step is to identify the task the operator has to perform to carry out the actions indicated on the Operator Action Event Trees (OAET). The~second step involves determining the instrumentation and controls required to carry out the identified tasks. The second step is carried cut independent of what is actually available in the control room.
Task Analysis Worksheet (BRP-CRDR-llA) Fields:
, 1. ' Event - The Event field is taken from the OAET. A typical Event name would be 'Small LOCA.*
- 2. Dominant Accident Sequence - This field is taken from the Dominant Accident Sequence Sets. These sets were developed from the PRA and the event trees that resulted from the PRA. -
- 3. '
OAET - The OAET identifier will be the unique identification used on the specific OAET bwing analyzed. 4. i Sequence Number s - The Sequence Numbers are the numbers assigned to each I logical step tue operator 90'uld take to mitigate an accident.
~
The
' steps coincide with the headings in the OAETs. The Sequence Number must be placed in the success or fail column under the sequence Number heading. The proper column can be. determined from the success / failure paths of the OAETs.
S. l Sequence Heading - Ths Heading used in the Dominant Accident Sequence i Sets should be entered in this column for the Operator Action groups covered by this Heading.
- 6. Operatoi Action - The Operator Action is a brief description of the task the operator is expected to do to mitigate an accident. Also equipment failure descriptions may be included here.
7. Action ^ Code - The action codes are indicated at the bottom of the form and are as follows: j -I- Information-the operator needs to make a decision to perform a task. A- Operator action required to control the system. F- Feedback to verify operator action was a success or failure. Doc ID 8760A/OS17A
r' BRP-CRDR-11 Page 2 of 2 8. Required Indicator or Action - The type of indication, control, or control action required to monitor or control the system should be entered in this column. Typical monitoring type entries would be - Pressure, Temperature, Level or Flow Trend or discrete indication, discrete alarm, system alarm, indicating lights, etc. Typical controls would be push buttons, control switches, selector switches, toggle switches, analog control station raise / lower buttons, setpoint adjustments, etc. Typical control actions would be close, open, raise, lower, trip, run, stop, automatic, adjust, etc. 9. Indication Position or Range - Typical entries in this column would be the process range required for the accident condition, equipment status lights, process limits, etc. 10. Scale Resolution - This refers to scales found on analog devices such as meters, dials and recorders. The required resolution should be indicated as a plus and minimum percentage of the required range. 11. Operator Response - This refers to the speed of response required by the operator in determing a parameter or taking an action. Typical entries are Rapid - Under time pressure Deliberate - Considered but timely action or reading required Analyze
- No time constraints to consider reading or action , 12.
Comments - Any notes required to clarify any of the various columns can be entered here. l i Doc ID 8760A/0517A
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BRP-CRDR-llA TASK ANALYSIS - TASK IDENTIFICATION WORKSHEET BY: VENT: DON. ACC. SEQ.: OAET: DATE: ISEQ NO. SEQ. BUC- HEAD- ACTION REQ'D IND. INDICATION SCALE OPERAMR
- ESS FAIL ING OPERAMR ACTION CODE OR ACTION POSITION OR RANCE RES. RESPONSE CONMENTS
- TION CODES: I - INFORMATION OPERATOR NEEDS M MAKE DECISION M PERFORN TASK A - OPERATOR ACTION REQUIRED TO CONTROL THE SYSTEM F - FEEDBACK TO VERIFY OPERATOR ACTION NAS A SUCCESS OR FAILURE pc 10 8760A/0517A
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ilRP-CRDH-1IA TASK ANAL,YSIS - TASK IDENTIFICATION WORKSifEET BY: 1 VENT: *qEO'-* OAET: DATE-EFQ NO. SEQ. SUC- HEAD- ACTION REQ'D IND. INDICATION SCALE OPERA'!UR CESS Fall, ING OPERATOR ACTION CODE OR ACTION POSITION OR RANGE RES. RESPONSE COMMENTS f>c_TCiA[,f eps os.re,ve a ii= :c X BY, Vos et~ r -d im O flu,~. Trof
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DHP-CHDR-11A TASK ANALYSIS - TASK IDENTIFICATION WORKSIIEET EVENT: BY: SEQ.: OAET: DATE: SEQ NO. SEQ. SUC- HEAD- ACTION REQ'D IND. INDICATION SCALE OPERA ~IUR CESS FAIL ING OPERATOR ACTION CODE OR ACTION POSITION OR RANGE RES. RESPONSE COMMENTS
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F t i l l l i i ATTACHMENT F Validation - Walkthrough (Procedures and Sample Documentation ll l . h
I BFS-CRDR-14 VALIDATION-WALKTHROUGH INTRODUCTION: - 7
&T)S 1 + kl A L /2 / ,, For the purpose of this procedure, the following definitions apply:
- 1. Walkthrough - Method of task analysis whereby control room operators conduct a step-by-step enactment of their actions during a transient scenario for an observer review team without actually carrying out the control function.
- 2. Validation - The process of determining whether the control room operating crew can perform their tasks effectively given the control room instrumentation and controls, validation implies a dynamic performance evaluation.
- 3. Dominant Accident sequence - A scenario comprised of two or more l
sequence headings in succession where the postulated successes and failures have a 10 per year or greater probability of occurrence. t All dominant accident sequences progress eventually to core melt by I'
, design. The Dominant Accident Sequences are from the Big Rock Point Probabilistic Risk Assessment.
l T ~, Il Doc ID 9290A 711A l l
O t BRP-CRDR-14 Page 2 1 DISCUSSION: Structured walk-throughs shall be conducted in the BRP control room. Structuring of the walk-throughs will insure that task. analysis can be validated and instrumentation identified through verification is satisfactory. Also consistent data from one sequence to the next can be i collected via question and answer sets that are uniformly applicable. Format for any particular walk-through will be: .
- l. Brief The Operating Crew
['. i:
;, Team members will provide the crew with a document which identifies 4
r, the sequence and describes the flow of the sequence from the initiators through each sequence heading. Once the intent is
; understood the team and crew will adjourn to the control room to i
conduct the walk-through. I'
,. 2. Entry Conditions l
L; l ~' Crew members will be informed of the initial plant states and their i present condition. Each instrument utilized by the crew to confirm these conditions will be identified by them and recorded by team members. Doc ID 9290A 711A
i BRP-CRDR-14 Page 3
- 3. First Sequence Heading t
Crews will be advised as to the plant states and equipment failures associated with.the first sequence heading. They will then be expected to consult procedures and identify actions associated with j the steps in the procedure. It will be permissible to redirect the operator to the expected procedures and actions if the identified actions are other than those dictated by the accident scenario. Actual initial operator actions will be recorded, however, for i future evaluation as to identification of potential HED's and for feedback to the task analysis.
- 4. Operator Action i.
After completing the simulated manipulation of controls dictated by
, their procedures, the crew will identify (either after or
{- throughout the operator actions) which instruments and controls were used to complete the task. Once completed, the crew will be advised of equipment failures and new pertinent plant states. Crews will then be instructed as to which sequence heading is next and asked to identify which instrument (s) indicates he has to take the action associated with that heading. Doc ID 9290A 711A
BRP-CROR-14 Page 4
- 5. Sequence Progression Step 4 shall be repeated until all sequence headings have been completed for the particular scenario being evaluation.
,! 6. Debriefing Team members and crew will be assembled to recap the scenario and i complete the documentation if necessary.
Scenarios to be covered will include those recommended in NUREG 700 and supplemented by additional scenarios. These additional scenarios are required i to cover all identified sequence headings from task analysis to assure us that
- all control room work stations are analyzed.
Scenarios (Sequences) are as follows: . FROM N" REG 700
- 1. Plant Startup
- 2. Plant Shutdown.
! Doc ID 9290A 711A
. . - . -~ . _ . _ _ . - - . . . . _ _ _
r . BE?-CR3R-14 Page 5 i Numbers 3 through 5 a11' represent inadequate core cooling'as the failure of each sequence leads to core damage. t
- 3. LOCA and Small Steam Line Break.
f , g
! 4. ATWS.
- 5. Spurious Reactor Depressurization System Operation.
l SUPPLEMENTAL SCENARIO'S TO COVER ALL OTHER READINGS AND ASSOCIATED WORK STATIONS
;, i
- 6. Loss of main condenser.
- 7. Turbine Trip (Reactor SCRAM).
- 8. Loss of Air.
- 9. Station Blackout.
I,. 10. Loss of DC Power. i
- 11. Spurious opening of the Turbine Bypass valve.
Doc ID 9290A 711A t
BRP-CRDR-14 Page 6
- 12. Figh Energy Line Break disabling normal core spray.
Personnel which developed the written task analysis shall be utilized as review team members. Operations shall provide and schedule control room l' operating crews to walk through the various scenarios. It is intended that as many crews as possible will be utilized for walk-throughs. For the purpose of BRP-CRDR a crew is defined as all control room members of a normal shift less the auxiliary operators. Due to the confined space within the control room, it has been determined that th,e reduced crew and team members will be ( sufficient to perform walk-throughs without overcrowding conditions l occurring. Walk-throughs shall be conducted while the unit is on-line and I hence no controis shall be manipulated and when requested, the team shall evacuate the control room. a f Data to be collected will be sufficient to validate the task analysis and provide operator / plant, operator / procedure and procedure / plant interface information. The latter will be obtained through a structured set of questions. All data collected will be used to generate HED's as necessary. Doc ID 9290W 711A i (
-nn ,,,, - ,,n, , - ,, , , - - , - - - - , ~ , - - - - - . , - - - - - , - - - - - - , - - - - , - - - . - - -- -
=
a BRP-CRDR-14
'Page 7 FORMS: ,
BRP-CRDR-14A VALIDATION /WAL'.-TEROUGH This form shall be used for each walk-0through and each scenario shall be
~
! identified with a unique supplemental number. (i.e. Plant Startup BRP-CRDR-14Al, LOCA BRP-CRDR-14A2, etc) . A cover sheet briefly describing the events in the scenario.shall accompany each package. Then, the data recording form shall contain preprinted information such as sequence, sequence heading,
~
operator action, required indication and' operator response from the task analysis and expected instrument number from the control room inventory. Areas to be_ filled out during the walk-throughs will be actual instrument used and comments. f 1 BRP-CRDR-14B QUESTIONS AND ANSWERS This form shall be used to record operator and review team responses to questions. These will be designed to explore the operator / plant, operator / procedure, plant / procedure and training interfaces. Some questions will be designed to provide information'on a sequence heading basis while s others shall be answered when an entire dominant accident sequence is completed. Doc ID 9290A 711A
?-
BRP-CRDR-14 Page 8 BRP ~RDR-14C CONTROL ROOM MOVEMENTS / WALK-THROUGHS This form shall be used to record operator movements while performing specific operator actions or basic movements to accomplish an entire sequence in I i general, f
- 1.
i - 1
-r Doc ID 9290A 711A
i i BRP-CRDR SEQUENCE-SCENARIO FOR SMALL LOCA SEQUENCE HEADINGS: S , RPS, G, I, Ycf, Ev, Cr, Cs, Lp 3 This scenario is initiated by a small steam line break inside containment (S3 ). Due to operator observations, the reactor is manually scramed. Also,
, due to operator observations, the enclosure spray system is manually placed in service prior to automatic actuation and is observed to be functioning.
c3eSRY*t-Isolation of the primary system has taken place due to outometeo % of the MSIV or the turbine stop or control valve is closed and the turbine bypass 4 and turbine bypass isolation valves are closed. It is observed that the feedwater system is supplying water to the primary system. The emergency r
, condenser was manually placed in service and observed to be functioning. The RDS was manually actuated in order to put the post incident system on recycle. Core spray was placed in service by successfully establishing a path
'l j for core spray and supplying adequate water flow from the fire pump. At some later time the post incident system has failed due to core spray pump failure, core spray heat exchanger failure, one train of core spray valves f ails to open, or the fire pump f ails. l l. 4 ,i Doc ID 9337A 713A
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F ATTACHMENT G t Control Room Inventory (Procedures and Sample Results) 9 f
1.EGEIS Control 1G5ii"Tnventory Inst No - Instrimment Essbers to be listed alphabetically; te, PS 101, PS 102, etc. System - System designation to be three-letter code; te, AES, BLS, CAS, etc. Service - Abbreviated neee of device; te, for F1 368, Service column would read "FD WTR HP HTR OUTLET" Panel tocation - Drawing number on which physical location of instrissent to shcun; te, C40165, 640166, etc. Ranae or Set Paint - Range and act points in appropriate values; te, "0-400 inches H20" "0-700 pst" or "1 RIP AT 100 pst RISTNG, RESET PflN." Interconnections - Interconnected instrissentation (pneumatic and electrical). T y - Instrumentation type: B-Bubbler, C-Capillary, D-Direct, E-Electric, F-Float, P-Pneumatic, R-Rotameter, TC-Thermocouple. Scale Itesolution - Readability lised - Y = Tes; N = No Rf: VISION DATE Blank 11/22/85 l 1 i l i I OTWTROI.-T W 4
-~ -- -- __ , , .
Cort 1tOL INVENTUR Instrument No System Service Panet location Range / Set Point Interconnection g Scale Resolution Used nee A-11b GS Condensate Col-A On/Off 52-12/CS E Red Light Pump Close,No d 1 A-11e CDS Condensate C01-A On/Off 52-12/CS E Creen Pump No 1 Light Tripped A-12h CDS Condensate C01-A On/Off 52-22/CS E Red Light Ptep No 2 Closed A-17e CDS Condensate C01-A On/Off 52-22/CS E Creen Pump No 2 Light Tripped A-16b CWS Cond Ctre Wtr Col-A On/Off 52-21/CS E Red Light Pump No 1 Closed A-16e CWS Cond Ctre Wtr C01-A On/Off 52-21/CS E Creen Pump No 1 Light Tripped A-17b dis Cond Ctre Wtr C01-A On/Off 52-11/CS E Red Llaht Ptap No 2 Closed A-17c CWS Cond Circ Wtr Col-A On/off 52-11/CS E Creen Pump No 2 Light Tripped A-IRb ECS Sta to Emerg Col-A On/Off RMC-5503 E Red Llaht Cond Valve MO-7062 MO-7062 Open A-Inc ECS Stu to Emerg C01-A On/Off RMC-5503 E Creen Cond Valve MD-7062 Light MO-7062 Closed A-19b ECS Cond From Emerg Col-A On/Off RMC-5504 E Red Light Cond Valve MD-7063 MO-7063 Open A-19e ECS Cond From Emerg Col-A On/Off RMC-5504 E Creen Cond valve MD-7063 Light MO-7063 Closed A-70h ECS Sta to Emerg C01-A On/Off RMC-5507 E Red Light Cond Valve MO-7052 Mn-7052 open 1 CONTROL LIST 11/25/1985
OW11tCL RonM INVENTURY Instruent No System Service Range / Set Point Interconnection Panet location M Scale Resolution Used M AM-114 FWS Reactor Feed Col-A Scale: P85 E *5 Amp Pump No 2 Amp 0-500, 500/5 AM-115 CDS Cond Pump Col-A Scale P9A E 15 Amp No 1 Ammeter 0-400, 400/5 , AM-116 CDS Cond Pump Col-A Scale: P95 E 15 Amp No 2 Ammeter 0-400, 400/5 AM-117 CWS Cond Circ Wtr C01-A Scale: P10A E *5 AMP Pump Mo 1 Amp 0-400, 400/5 i t AM-118 CWS Cond Cire Wtr Col-A Scale: P10B E 15 AMP Pump No 2 Amp O-400, 400/5 FI-2330 FWS Feed Wtr to C01-A 0-300 lb/hr FT-2160 P *10 lb/hr Sta Drum x 1000 Img x 1000 i FI-2333 PIS Enclosure Sprays Col-A 0-400 CPM FT-2164 E *2-1/2 CPM
~
Los F1-2334 PIS Backup Enclosure C01-A 0-400 CPM FT-2161 E 12-1/2 CPM Sprays tog .I FI-2335 PIS Core Spray 001-A 0-900 CPM FT-2162 E *25 CPM l FI-2336 PIS Redimdant Core C01-A 0-900 CPM FT-2163 E 125 CPM Spray . . log HS-7000 PWS React Fd Pump Col-A Open SV-4852 E NA No 1 Rectre CV-4001+ A-6b j HS-7001 FWS React Fd Pump Col-A Open CV-4002 E NA (A-7a) No 2 Rectre SV-4853 A-7b Its-7003 CIS Reactor Enc C01-A Close/Open CV4031/4102 E NA Clean Sump Disch PR-7501-1 1 IET-7501-1 l a wrRot. LIST 11/25/1985 6
I . i I ATTACHMENT H Task Analysis - Instrument and Control Verification Worksheets i i a
__ --- . . . . . . , ,, ' ~~' YEZIFICOTIosynueWLSHks e TASK ANALYS13 - INSTRUMENT AND CONT SUITABILITY AVAILABILITY ._ Y ASJLANALYSIS REQUIREMMTS DIr EES Range S/oc Scale ta h Scale Irmtramient " System ties b bidsline C Seq No Required No Parwl 10 Service _ Set Point - Etzelullen ffeed. _ Character 1stlee _ __Resolut!an il PSIO _ _ PR-55 PIS Roset 81dg Press -5 to StsSS Press 11 pel CO2-1 115 PSIG 1.0 0-40 pelg (4-20 mal Reect Side Prese -5 to 11 PSIG _ _ PR-M PIS
$1855 CO2-1 115 PSIG 1.0 14-20 mal $1855 2.0 Steen Flow 0-100% 12.5% = _ _
FR-1006 FHS 25.000
$1855 Flow 150 100%=1H CO2-1 2.0 0-1046I lb/hr lb/h 83-153 lb/hr Feedseter Flou e-100% 12.5% = _ _
FR-ID07 FMS 25,000 , SlaSS CO2-1 100%=1H 2.0 lb/h 43-159 lb/hr
~
21 Hone S1855 tewel 2.0 0-40 Ft . 25 Hone S1855 Heisture 2.0 0-100% RN Hormel Range 0-5 V Verlebte _ _ RR-8057 SCH
$1855 Particulate 210 10-101Fl Log CO2-4 2.9 0-1000 cps VE;IF1 01/14/1986 i
I i i I
TASE ANALYSIS - INSTRLMENT AMB CONTROL' VERIFICATION SMEET T A!,K ANALYSIS REQUIREMENTS AVAILABILITY SUITABILITY l Seq No Regsired Scale Ins trtment System Range s/or Scale Die EES
],)sted. __ Characteri s tics _Renalutian_ No Panal.19 Servic L ,,.$et Paint Etzelullen fles gua Culdallne Ceeeante BED Ble ;S1335 level 13 In LR-IO12 FMS Steen Drue Level -25"-0*25" 2.5" _ ,,,,,,,
1.0 0-55 ft. CO2-1 , , ,. 51855 Pre m re 119 pel PR-1A09 FMS peset Prese 0-100% = 22.5% = _ ,,,,,,,_ , 1. 0 0-1750 pela CO2-1 0-2000 PSI 150 PSI f
$1453 Po see 25% RR-RIO5A Dets Fluw Level. 0-125% 11% .,,,,,,, _
1.0 101-69-154% CO2-1 . Cheemm1 3 , .. 9-18 HV, , , , . , , !Slass RR-RIO55 9995 Fluw Level. 0-125% 11% _ _
- 1.o CO2-1 Charme! 2 e-lo ew
)53355 RR-RIO5C DetS Fluw Levele 5-125% 11% _ ,,,,,,,,,
j 1.0 CO2-1 Charmel 1 0-10 IW . SisSS Level it it LR-3111 PIS Roset Bldy Level 579 Ft 2.2 Ft ,,,,,,, _ 1.0 0-40 ft. CO2-1 596 Ft '.51435 LR-3110 PIS Reect Blds Level SM Ft 1.2 Ft _ ,_, i 1.0 r CO2-1 596 Ft - 1 i St a*.5 T em. 15' TR-%23 VAS Centelneant Test s'-120'r il _ ,_, i 1.9 50-500*F CO2-4 Dem Pt IFembero NII 0 l 1 fVEE!F1 01/14/1966 SS-l l
r-ATTACHMENT I I Sample NUREG-0700 Annunciator Survey Results i I e h 9 P t
7 ANNUNCIATOR SURVEY (ANS) This survey covers items that are important determinants of the operation of annunciator systems. Examine each annunciator panel. ANS-1 Are tiles grouped functionally? l ANS-2 If alarms are prioritized, is the method of prioritization followed consistently? 1. ANS-3 List tiles with the following traits: . (a) They employ multiple-choice indication. (b) They have legends that do not unambiguously specify alarmed point or disagree with abbreviation / acronym list. (c) They are not associated with controls and displays on same panel segment. t
, ANS-4 List tiles that normally or frequently are on during nor- ,' mal operation.
O I 4 i
- 1. .
ig 4 A-71
ANNUNCI ATOR WARNING SYSTEMS 6.3 GENERAL SYSTEM CHARACTERISTICS 6.3.1 GUIDELINE COMPLIANCE CHECKLIST N/A Yes No Reference / Comment 6.3.1.1 GENER AL SYSTEM DESIGN . _
" ^ ~
Annunciator warning systems are the primary control room interface to immediately alert the , E ,
^-
operator to out of tolerance changes in plant con- . dition. Annunciator warning systems consist of . . _ ,
- c
.! three major subsystems: (a) an auditory alert , - -
f _ ""
s, .
. subsystem, (b) a visual alarm subsystem, and (c) an ., .
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~
operator response subsystem (see Exhibit 6.31). T ,. ^ j^ ,(
^
Together, these three subsystems should be designed l' + . . . . .9" ,s 7 ~ ,i to provide a preferred operational sequence for f annunciator warnings as indicated in Exhibit 6.3 2.
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ANNUNCIATOR WARNING SYSTEMS 6.3 GENERAL SYSTEM CHARACTERISTICS 6.3.1
/*
AUDITORY ALERT SUBSYSTEM f VISUAL ALARM SUBSYSTEM g.d, OPERATOR RESPONSE SUBSYSTEM
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Exhibit 6.3 1. Annunciator waming system. l - 1 f 6.3 2
e i ANNUNCIATOR WARNING SYSTEMS 6.3 GENERAL SYSTEM CHARACTERISTICS 6.3.1 Pleat Anerva Powasumes " ,""" Carerte a__g,
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I Exhibit 6.3 2. Annunciator system preferred operational sequence. 6.3 3
ANNUNCI ATOR WARNING SYSTEMS 6.3 GENERAL SYSTEM CHARACTERISTICS 6.3.1 GUIDELINE COMPLIANCE CHECKLIST NIA Yes l No Reference / Comment l 6.3.1.2 ALARM PAR AMETER SELECTION . ' Plant parameters selected for inclusion in the . a annunciator warning system and the limits or alarm
. ;j set points for those parameters should be established 9 to ensure compliance with technical specifications
( y% < ti l and to allow the operator to monitor the status of ~l the plant and respond to out of tolerance conditions 6 _ I effectively. The following general guidelines are suggested to increase the operational effectiveness of N ,
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the annunciator warning system. -
- a. SET POINTS-The limits or set points for initiating the annunciator warning system should be established to meet the following ;
j goals: u (1) Alarms should not occur so frequently ! as to be considered a nuisana by the . ., operators. '\ , (2) However, set points should be established j j
, to give operators pdequate tiruGo re- ,,
spond to the warning condition before a l serious problem develops.
- b. GENERAL ALARMS (
n ., I (1) Alarms that require the control room I , J p operator to direct an auxiliary operator '- "
, . J .' ,
I - to a given plant location for specific . .- ', y. I
- information should be avoided. ,
(2) If general alarms must be used, they i should only be used for conditions that i., - ' - I i ~. allow adequate time for auxiliary operator , action and subsequent control room , . , operator action. ! i c.) MULTICHANNEL OR SHARED ALARMS (, y., , -
- . (1) Annunciators with inputs from more than ^'
one plant parameter set point should h
/
- be avoided. Multi input alarms that summarize single. input annunciators 4 ,# '-
elsewhere in the control room are an j } q 1 g; exception. l. ii\ i 9 6.3 4
ANNUNCIATOR WARNING SYSTEMS 6.3 GENERAL SYSTEM CHARACTERISTICS 6.3.1 COMPLIANCE CHECKLIST N/A Yes No Reference / Comment 6.3.1.2 ALARM PARAMETER SELECTION (Cont'd) l
- c. MULTICHANNEL OR SHARED ALARMS i U ' i -- ( ~
(Cont'd) (2) Where multiinput annunciators must be . used, an alarm printout capability should ' D #' '
"'*2' jj be provided. The specifics of the alarm . . . .
should be printed on an alarm typer with i l' sufficient speed and buffer storage to II .e v.
'll capture all alarm data.
(3) A reflash capability should be provided to l l ,. . -- 's'- .,. , allow subsequent alarms to activate theh j '
' ' " ~
auditory alert mechanism and reflash the
,I ,
-- *'~- -
l visual tile even though the first alarm ma m
' * ' I not have been cleared. -
- A ' ' t' l l
- d. MULTI UNIT ALARMS-Multiple-unit nuclear
, up y y , W 4
[ ' power plants may contain plant systems that N
, a
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4 are shared by two or more units.The status of , _
. any such equipment should be provided in all . W $9
" ' " I control rooms.
(1) Alarms for any shared plant systems - N should be duplicated in all control rooms. ' (2) When an item of shared equipment is being operated from one control room, l i a status display or signal should be pro- ' vided in all other control rooms which l could potentially control this equipment. i f i f a 6.3 5
O ANNUNCIATOR WARNING SYSTEMS 6.3 GENERAL SYSTEM CHARACTERISTICS 6.3.1 GUIDELINE COMPLI ANCE CHECKLIST , N/A Yes No Referones/ Comment 6.3.1.3 FIRST OUT ANNUNCIATORS . m
+
- x nw '
As an aid to diagnostic procedures, provision should be made for identifying the initiating event (first d ;
. , , f' ,
4
- e. m .
^
out) associated .with automatic plant shutdowns. s . s I
- a. REACTOR SYSTEM
{ (1) A separate first out panel should be provided for the reactor system. h 7
/
[ "l (2) The first out panel should consist of [ ' separate annunciator tiles for each of the - i i
' automatic reactor trip functions. ! !
i [ (3) In the event of a reactor trip, the tile I
! i associated with the event should j j illuminate, and no other. i
- b. TURBINE GENERATOR SYSTEM-A separate first out panel, similar in function to the/
h '! i reactor system panel, is recommended. i I
.\ \, ~-
POSITION-First out panels should be located j c.
# directly above the main control work station I","'
h d. for the system. APPLICATION-First out annunciators should \
't ,/-
'I :, .
I. conform to the general auditory, visual, and l. <,.,.- c- { operator response guidelines of this section, i i 4 1* ' - 1 t .
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6.36 i
l ANNUNCIATOR WARNING SYSTEMS 6.3 GENERAL SYSTEM CHARACTERISTICS 6.3.1 GUIDELINE COMPLIANCE CHECKLIST N/A Yes No Reference / Comment 6.3.1.4 PRIORITIZATION a , Because of the large number of annunciators e.z ,. n8 ;
- typically found in control rooms and the likelihood ? ;4 +
that numerous alarms may come in concurrently, i ,4 ;_ some logical prioritization should be applied such ;.,3 fjj s. . ga ~ m; ' ' " that operators can differentiate the most important j? . # fG ^ , <
+
or serious alarms from less important ones. w : e - W? l
- a. LEVELS OF PRIORITY g g g (1) Prioritization should be accomplished l
/ C 'ffT using a relatively small (2-4) number of i
# ,, priority levels. (
i (2) Prioritization should be based on a continuum of importance, severity, or j need for operator action in one or more i dimensions, e.g., likelihood of reactor trip, release of radiation. Exhibit 6.3-3 ~ provides an example of prioritization f
.I based on three levels of prioritization. i FIRST PRIORITY ALARMS e Plant shut down (reactor trip. turbine trip) ,
o Mediation release g7 ,- ,,.
, g ,
o Plant conditions which. if not corrected immediately. vdll result in automatic plant shutdown or redietton e roleses, or will reevice menuel plant shutdown. k, , l' ". !I SEcoNo PRIORITY ALARMS e Technical specification violations which if not corrected l vnll roovire plant shutdown I e Plant conditions which. If not corrected. mey lead to plant shut down or radiation releases i THIRD PRIORITY ALARMS e Plent conditions representing problems (e.0.. system i ! degredation) which effect plant operstplity but which should not lead to plant shutdown, radiation l reisene. or violation of technical specifications e Exhibit 6.3 3. Three level annunciator prioritization example. 6.3 7 l
e44 r 4 i f ANNUNCIATOR WARNING SYSTEMS 6.3 GENERAL SYSTEM CHARACTERISTICS 6.3.1 COMPLIANCE CHECKLIST 6.3.1.4 PRIORITIZATION (Cont'd) i N/A Yes l No Referones/ Comment
- b. PRIORITY CODING I l (1) Some method for coding the visual signals ,
- for the various -priority levels should \i ,
I be employed. Acceptable methods for - i priority coding include color, position, j shape, or symbolic coding. ! (2) Auditory signal coding for priority level is , l also appropriate.See Guideline 6.6.1.4 for \ t recommended coding techniques. 9 I
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f I w 4 t 6.3-8 y
ANNUNCIATOR WARNING SYSTEMS 6.3 GENERAL SYSTEM CHARACTERISTICS 6.3.1 GUIDELINE COMPLI ANCE CHECKLIST N/A YeslNol Refmnon/ Comment 6.3.1.5 CLEARED ALARMS .. j
' J.,
Alarms that have cleared should initiate an audible *
'T ,
signal and activate the alarmed and acknowledged , ', ^ visual tile in some way to provide positive con- , '7 '
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T~ firmation. p - c .,,, . i , y
- a. AUDITORY SIGNAL-Cleared alarms should .
,q have a dedicated, distinctive audible signal j '\, . . y i..,,
U,s ^ 0-w W ' g,
; f/ which should be of finite duration.
- b. VISUAL SIGNAL-The individual tile should
~
l (.P have one of the following: . (1) A special flash rate (twice or one-half the ; normal flash rate is preferred, to allow -(' i ?
! l discrimination), or j MI (2) Reduced brightness, or (3) A special color, consistent with the overall control room color coding scheme, pro- .
j duced by a differently colored bulb ,
,' behind the tile.
\ \ I r 6.3 9
ANNUNCIATOR WARNING SYSTEMS 6.3 AUDITORY ALERT SUBSYSTEM 6.3.2 GUIDELINE COMPLIANCE CHECKLIST N/A Yes l No Nfmace/ Comment l 6.3.2.1 SIGNAL DETECTION .w .
, y. 4 ,
The first requirement for an effective annunciator warning system is rapid penetration of the [h[ , [,.; s y -
' if . _ ,
operator's awareness under any and all conditions. .
~
j e - Some " worst case" conditions might be: (a) during m
^
% p. c .,
! the first graveyard shift after four days off duty, %
% S%sM'[ ' f';
(b) the sixteenth hour of work occasioned by % - D ' ~ ,~ i the sudden illness of the relief operator, (c) a warning signal appearing while the operator is very
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busy attending to the three or more immediately
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preceding problems, or (d) a warning signal .
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appearing after a series of false or nuisance alarms. , I,~ Specific principles for the auditory alert subsystem 4 ;,
~>
follow. . 11 l* a. INTENSITY-The signal should be such that operators can reliably discern the signal above the ambient control room noise. A nominal value of 10 dB(A) above average ambient noise is generally adequate.
}
g b. CONTROL -Signal intensity, if adjustable, g
- A: , ., ,
J
/
Q should be controlled by administrative ls '
/ procedure. !
,.S- s #' -[
I c. LIMITS-The signal should capture the l i operator's attention but should not cause ,,; j irritation or a startled reaction. 1 i
- d. DETECTION-Esch auditory signal should be ! ,
l adjusted to result in approximately equal !' , l detection levels at normal operator work j
! stations in the primary operating area. : !
- e. RESET - The annunciator auditory alert ,
mechanism should automatically reset when it l
! has been silenced. j l l .
- f. IDENTIFICATION-The operator should be
; able to identify the work station or the system j where the auditory alert signal originated. ,
Separate auditory signals at each work station i within the primary operating area are i I recommended. l 6.3-10
. - . _ . - - ~ . . .- _. . - - _ - __ _ -. ._ - -. . . _.
0 ANNUNCIATOR WARNING SYSTEMS 6.3 AUDITORY ALERT SUBSYSTEM 6.3.2 GUIDELINE COMPLIANCE CHECKLIST N/A Yes No Reference / Comment l 6.3.2.2 CODING . 7 -
, - i u - <* '
i Auditory coding techniques can improve operator ' ~ M $' a't . 1 n:-5;^
. response to auditory alarms. This is an important ,, ,,
~~
consideration when auditory alarms are not 2 ...g r-accompanied by a visual alarm in the primary ,, [y l$ ~~ l (I;ff[2. ' , t-operating area. Auditory coding techniques are % '3IE M , . , f 'W 9 , presented in Guideline 6.2.2.3. A + b' W E@ "M -
- a. LOCALIZATION . . Q. g. [. i: *.,
i 1 (1) Auditory coding techniques should be ,4 .e.- m c.J . i f f. used when the operator work station : associated with the alarm is not in the j ' g r. ,.. - ~. , l primary operating area. ! -
'l_._ i (2) Coded signals from a single audio source should not be used to identify individual A work stations within the primary oper-ating area.
- b. PRIORITIZATION-Coding may be used to [ g i indicate alarm priority. (See Guideline 6.3.1.4.) i D -
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e 1 L'
+.
l l 6.3 11
ANNUNCIATOR WARNING SYSTEMS 6.3 VISUAL ALARM SUBSYSTEMS 6.3.3 l l GUIDELINE COMPLIANCE CHECKLIST , N/A Yes l No l Reference / Comment l 1 6.3.3.1 VISUAL ANNUNCIATOR PANELS : Annunciator panels are generally organized as matrices of visual alarm tiles. These matrices, separated into groups by function, are usually ' positioned along the top of the control boards or ,, ; , > consoles (see Guidelines 6.1.2.2 and 6.1.2.3).
- a. LOCATION-Visual alarm panels should be lIn\ , . M ) ; ._ _ . . , ,l I located above the related controls and displays -
,"i
- which are required for corrective or diagnostic , Ok j' . " i'" ~ ' ~* f.f' J '- '-
'i ,
Co action in response to the alarm. (See Exhibit- - - ' , '- l .
- / - 6.3 4.) l s . a ,i m
- b. LABELING .-
- f. - ,
(1) Each panel should be identified by a label /
! above the panel. ,
4 I li',
'j'
(2) Panel identification label height should be l l l consistent with a subtended visual angle i ; l of at least 15 minutes when viewed from a \, { central position within the primary
, operating area.
- c. LAMP REPLACEMENT : ,
(1) If a lamp replacement requires legend tile q; , removal, there should be a way to ensure i that the tile is replaced in the correct location. (2) Lamp replacement should not subject the . operator to a shock hazard. .' r (3) Operator aids should be provided if . needed for lamp replacement. 6.3-12
i l ANNUNCIATOR WARNING SYSTEMS 6.3 VISUAL ALARM SUBSYSTEMS 6.3.3 VISUAL eseo.aten contnot ==' "" AuRMS 8 . j .
, . ,1 l
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l RELATED 1, :D i 1 1 I I ' 1 ,1 1 1 1 =>t
. [ DISPLAYS as e W
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on ammmmmmmad a C M e~ M eC M a w""
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~
-sa RELATED m.
i . .. --l . ._ i III 10 th W m
- .. 3 CONTROLS t:I g _ " t3 p s
t:1 t*l ca W 2 --- v 8... g_= b j = em em ~ g i td m L~ -. 3~, ( l td m 5~, f
, a j
. * *_ Z .
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~
I m - e 1 U [LD ' Q l , . lIl O m v l td 6 : J5 t Exhibit 6.3 4. Visual alarms located above the related controls and displays. (From Seminara et al.,1979). 6.3-13
ANNUNCIATOR WARNING SYSTEMS 6.3 VISUAL ALARM SUBSYSTEMS 6.3.3 GUIDELINE COMPLI ANCE CHECKLIST N/A Yes No Reference / Comment l
, w . . -.. -
.w 6.3.3.2 VISU AL ALARM RECOGNITION AND IDENTIFICATION
; Specific guidelines for visual recognition and identi- -
fication of alarmed tiles follow: ' I
- a. FLASHING - The specific tile (s) on an annun- ,
l ciator panel should use flashing illumination to (, . I ! Indicate an alarm condition. d ,,..,,'. d f
! b. FLASH RATE - Flash rates should be from da i
,I\ h*'
#'f, three to five flashes per second with approxi-u ,.i_-
f[/ mately equal on and off times. . '.'!u.c.
- c. FLASHER FAILURE - In case of flasher '
I(/ h' '7@, h p Y'. ." ' ' M/. failure of an alarmed tile, the tile light should ih 44' M .d4 M N illuminate and burn steadily. ' [ .
- d. CONTRAST DETECTABILITY-There should . :
be high enough contrast between alarming and steady on tiles, and between illuminated and k I i non illuminated tiles, so that operators in a ' normally illuminated control room have no problem discriminating alarming, steady-on, and steady off visual tiles. '
" DARK" ANNUNCIATOR PANELS- A e.
" dark" annunciator panel concept should be .
l used. This means that under normal operating 9i l . 1 i conditions no annunciators would be illumi- ! nated; all of the visual tiles of the annunciator ! l i i panels would be " dark." EXTENDED DURATION ILLUMINATION-
- f. l+
If an annunciator tile must be "ON" for an d.{. .- . extended period during normal operations '
^ ' " # '
" ' + ,
(e.g., during equipment repair or replacement), it should be: l L * ..; ,. ,I l i (1) Distinctively coded for positive recog- i nition during this period, and t
! IV % ' >=. .
(2) Controlled by administrative procedures. l ' I N . l l 6.3-14
D
- l. N . 4 A. - b_ Lk' l' N ANNUNCI ATOR WARNING SYSTEMS 6.3 VISUAL ALARM SUBSYSTEMS 6.3.3 ,
GUIDELINE COMPLIANCE CHECKLIST ; N/A Yes No Reference / Comment ,. 6.3.3.3 ARRANGEMENT OF VISUAL ALARM , p t p.3 p. l ,t
- a. MATRIX ORGANIZATION - Visual alarms u +-+ U w My
) /[%.T,,,, ,
should be organized as a matrix of visual alarm y. ( , / . . I . , <' ' . t'b, f( s
~
tiles w.th.in i each annunciator panel. x, . w. _ _ . l ..f t *
- b. FUNCTIONAL GROUPING-Visual alarm tiles 6 should be grouped by function or system ,
I within each annunciator panel. For example, area radiation alarms should be grouped on one Q _ panel, not spread throughout the control ,
. room. !
- c. LABELING OF AXES ,....
<.r., . ,, .
(1) The vertical and horizontal axes of . . .j ;
' ^
' s w : ' N 't -
/ annunciator panels should be labeled - -
with alphanumerics for ready coordinate designation of a particular visual tile. l h b'
,c " ' ' '
7
" -~ ~~'h' l
';^'<.
(2) Coordinate designation is preferred on the 7 'i left and top sides of the annunciator g j ( panel. \ t .c j, 'g. , r, g,f(
/-
j (3) Letter height for coordinate designation , should be consistent with a subtended , visual angle of at %st 15 minutes as Y ... . J.. 2. viewed from a central position within c, .
' w- A ' " ' 'I the primary operating area.
- d. PATTERN RECOGNITION -When a panel Y 4% M contains a number of simultaneously illumi- M $N M % %)@ W ^ [~ ^
*@%8 1 dh@ihd5[dij7 NM J ~
nated visual tiles, pattern recognition can help ' in diagnosing and mitigating problems. Pattern, W " M $N$$$2$db' recognition can be facilitated in annunciator
" W ,
panels by meeting the following criteria:
'M ' '
' ' Y./ "'
'I
! (1) The number of alarm tiles and the matrix .t 'd. .s
{ density should be kept low (a maximum q '[ 's j of 50 tiles per matrix is suggested). , ,
! (2) Tiles within an annunciator panel matrix up " - .
should be grouped by subsystem, function, I ,
', f,, u'f, ;
' , {,
J or other logical organization. ,,~.s~.
.... 7 ,i -
F 6.3-15
ANNUNCI ATOR WARNING SYSTEMS 6.3 VISUAL ALARM SUBSYSTEMS 6.3.3 COMPLIANCE CHECKLIST N/A Yes No Reference / Comment 6.3.3.3 ARRANGEMENT OF VISUAL ALARM TILES (Cont'd)
' ~
" ,1 -
- e. OUT OF SERVICE ALARMS-Cues for prompt -
{J t-' , ' I 3 W-
'. / - recognition of an out-of-service annunciator should be designed into the system. ! ! !' f. BLANK TILES-Blank or unused annunciator I I tiles should not be illuminated (except during annunciator testing).
Yj ' l t i t 6.3-16
0 *t % e' D W sf NY ~. ANNUNCI ATOR WARNING SYSTEMS 6.3 VISUAL ALARM SUBSYSTEMS 6.3.3 GUIDELINE COMPLIANCE CHECKLIST I N/A Yes !No Reference / Comment 6.3.3.4 VISUAL TILE LEGENDS
] _ , ., '
UNAMBIGUOUS - Annunciator visual tile - . p a. legends should be specific and unambiguous. Wording should be in concise, short messages. I A i Y'
~ ~ [ I T-'4['l' .3]
'l 3 l$
- b. SINGULARITY - Alarms which refer the .
operator to another, more detailed annunciator b panel located outside the primary operating area should be minimized.
! c. SPECIFICITY -Tile legends should address A ,,
6, i
/
f c. specific conditions; for example, do not use one alarm for HIGH LOW, TEMPERATURE-
/, { w '- ' , ' l; e l
l.
~
PRESSURE. s ,
- d. ABBREVIATIONS- Abbreviations and acro- -
f- b '.: 4 \.._ .8 .- nyms should be consistent with those used i / ( ;. F _., ,,, s
- elsewhere in the control room. M' '? r.->. -
[/ / , i p. . '. y, ( . , *.. e 6.3 17
ANNUNCIATOR WARNING SYSTEMS 6.3 VISUAL ALARM SUBSYSTEMS 6.3.3 GUIDELINE COMPLIANCE CHECKLIST N/A Yes No Refmnce/ Comment 6.3.3.5 VISUAL TILE READABILITY
- a. DISTANCE-The operator should be able to , 3 l
read all the annunciator tiles from the position j at the work station where the annunciator i l acknowledge control is located. ! l 4 (1) Letter height should subtend a minimum , visual angle of 15 minutes, or .004 x y viewing distance. The preferred visual i i angle is 20 minutes, or .006 x viewing j t distance. (2) Letter height should be identical for all i tiles, based on the maximum viewing ,- ! l distance. Separate calculations should be , made for stand-up and sit-down work stations.
- b. TYPE STYLE-The size and style of lettering i f j
. should meet the following: !s 1 I s (1) Type styles should be simple. i ! ;
(2) Type styles should be consistent or, all i i " lc t. % - .,V visual tiles. T . -< s - - -; [x f! (3) Only uppercase type should be used on '
, visual tiles. .' N I
[ c. LEGEND CONTRAST-Legends should pro- l ! vide high contrast with the tile background. l i (1) Legends should be engraved. -
! (2) Legends should be dark lettering on a if in <- 6 J \/
light background. < i
'C , f y604 I
[
)td'2p'*hI 1
a
;4d i
6.3-18 i
ANNUNCIATOR WARNING SYSTEMS 6.3 VISUAL ALARM SUBSYSTEMS 6.3.3 COMPLIANCE CHECKLIST N/A Yes l No Reference / Comment l 6.3.3.5 VISUAL TILE READABILITY (Cont'd) ,
, ,, 4
- d. LETTER DIMENSIONS AND SPACING - ^Jc:;Jf ;.
Recommended dimensional characteristics of
- 5 L ~', 9 ~;
'~'
legend lettering and spacing are as follows: ~ ' "I " e ~ '3A ' (1) Stroke width to character height ratio ,. should be between 1:6 and 1:8. 's ' (2) Letter width-to-height ratio should be , between 1:1 and 3:5. N (3) Numeral width-to-height ratio should be . , 3:5. l (4) Minimum space between characters should - l' be one stroke width. - (5) Minimum space between words should be ', j the width of one character. l (6) Minimum space between lines should be ! I one-half the character height. t l I e 6**
,.' I g
v.+
~.
l l l l 6.3-19
ANNUNCIATOR WARNING SYSTEM 6.3 OPERATOR RESPONSE SUBSYSTEM 6.3.4 l GUIDELINE COMPLIANCE CHECKLIST l N/A Yes No Reference / Comment 6.3.4.1 CONTROLS . w ~ Controls for operator response to the annunciator ^* system include silence, acknowledge, reset, and test . controls. (See Exhibit 6.3 5.) '-
- a. SILENCE (1) Each set of operator response controls ' .NO..,..,1>k.5-i should include a silence control. l l
, (2) It should be possible to silence an ,
I auditory alert signal from any set of , . . ","' annunciator response controls in the ls
.A p .~ ,- . .
[ primary operating area. ;
- b. ACKNOWLEDGE i !
(1) A control should be provided to terminate .i i the flashing of a visual tile and have it q' i , continue at steady illumination until the - alarm is cleared. - (2) Acknowledgement should be possible ,l only at the work station where the alarm 'i i originated. i ! j
- c. RESET (1) If an automatic cleared alarm feature is ,i i not provioed, a control should be pro- ;
vided to reset the system after an alarm ' has cleared. , i (2) The reset control should silence any , , audible signal indicating clearance and j) should extinguish tile illumination. : (3) The reset cor trol should be effective only f ; i at the work station for the annunciator : ' panel where the alarm initiated. ! , l
! d. TEST ;
(1) A control to test the auditory signal and flashing illumination of all tiles in a panel \ should be provided. l (2) Periodic testing of annunciators should be l l % p..s ) -. {i ( required and controlled by administrative ' j j j ., l j procedure. : l l 6.3 20 l
e ANNUNCIATOR WARNING SYSTEM 6.3 OPERATOR RESPONSE SUBSYSTEM 6.3.4 Example 1 rc O F 3 7 SILENCE ' ACKNOWLEDGE 9 k
~
[ C 7 E A I RESET TEST I -1 l Example 2
.. . . i .' . .$. '$. .U. .N.
. iiik:.:..t', 5. . .' .
. . d. .t'4. .T. O. N. .'.C. U.N.. Y. . R.6.
l SI ACK ' RESET , TECT l:::. . . ..... .:.:.:.:.: - k!!! !!! Exhibit 6.3 5. Annunciator response controls. l l' 6.3 21
ANNUNCIATOR WARNING SYSTEM 6.3 OPERATOR RESPONSE SUBSYSTEM 6.3.4 GUIDELINE COMPLIANCE CHECKLIST N/A Yes No Reference / Comment 6.3.4.2 CONTROL SET DESIGN _ i ,-
- a. POSITIONING OF REPETITIVE GROUPS- b b # h A.V Repetitive groups of annunciator controls , !! .
~
[ should have the same arrangement and relative h,f, j g,' y, 5 ' .g I * ,' "' g ,"9. location at different work stations. This i ' " '" ~' " ' # - ? ' ' 7# is to facilitate " blind" reaching. l 1.a . . . . 1 . '.- '
; b. CONTROL CODING- Annunciator response j I f
L' - controls should be coded for easy recognition :
! 'i '
k\'. ' { using techniques such as: (1) Color coding; , j i !
! i (2) color shading the group of annunciator 4/j !
controls; (3) demarcating the group of annunciator controls;or [j : (4) shape coding, particularly the silence 'q , control. (See Exhibit 6.3 5, Example 2.) .
/'
- c. NONDEFEATABLE CONTROLS- Annunci- l l ator control designs should not allow the operator to defeat the control. For example, 1
! >[ !l p g-[
some pushbuttons used for annunciator ;
' l silencing and acknowledgement can be held j
# down by inserting a coin in the ring around the i i pushbutton. This undesirable design feature , l' should be eliminated. '
9 l; 1, 1 6.3 22 l
l l ANNUNCIATOR WARNING SYSTEM 6.3 OPERATOR RESPONSE SUBSYSTEM 6.3.4 GUIDELINE COMPLIANCE CHECKLIST N/A Yes No Referenos/ Comment 6.3.4.3 ANNUNCIATOR RESPONSE PROCEDURES
- a. AVAILABILITY- Annunciator response pro- %, L ( e cedures should be wallable in the control ,
room.
- b. INDEXING-Annunciator response procedures .
I should be indexed by panel identification and % l
- annunciator tile coordinates.
4 l i
,P r
l lr i t 6.3 23 f l
l' ATTACHMENT J r i Human Engineering Discrepancy Documentation Procedures 1 and Forms i I L 4 f i
BRP-CRDR-7 Page 1 of 3 HUM AN ENGINEERING DISCREPANCIES This form addresses the recording, categorization and corrective action of discrepancies identified in any of the various surveys, inventories, checklists or any other document source utilized in the evaluation of the Control Room Design Review. Items identified which clearly or potentially violate human engineering principles will be recorded on this form. Categorization of each-identified HED is covered by a separate procedure. l Discrepancy Information 1.
- 1. HED Title An abbreviated description in general terms which can be used on the HED log sheet to describe and differentiate between the particular discrepancies.
From Review f' Title of survey, inventory or checklist utilized to initially identify the discrepancy along with the form number and sheet number if applicable. jl. 2. - Cross Referencing i !>- Item Number Either the actual instrument number or type of instrument. If the above is not applicable, enter the general area, an example being a discrepancy identified in the sound level surveys control room, containment, etc. I' Panel ID l If applicable, the control room panel number. System / Subsystem Enter abbreviation of the system / subsystem wherein the component is 7 utilized. Example: RCS, plant communications, ALP, etc. Modality Check the primary sense affected by the discrepancy. 4
^
BRP-CRDR-7 Page 2'of 3
- 3. Description Describe in sufficient detail the discrepancy identified. This description will be used in conjunction with classification information to properly categorize the HED.
- 4. NUREG 0700 ID i
Identify the applicable section and subsection of the NUREG (section 6) { which precipitated the HED.
- 5. Photo Log Number Record the appropriate photograph identification number. If the photograph used is for a generic situation, such as labeling or mimics, identify the number as generic.
Photograph Title An abbreviated description in general terms which can be used.on the { Photo Log Sheet to describe and differentiate between the various photographs. Classification Information
- 1. Access the potential operating error (s) that could result from the HED
- t.
and record the item (s) which could cause the error. l 2. List any known procedures or operator sequence where the cause of the error could lead to misoperation. l
- 3. List the consequence of the error.
Corrective Action 1. Provide a suggested solution and identify affected documents, training, procedure change, paint, tape, label requirements or backfits. Where applicable, note CRMP involvement and TRG interface required. e i s 2. Schedule will be completed based on the categorization process and recorded after the assignment by the appropriate review. General Comments Record any other pertinent data necessary to properly categorize the HED.
.n - - ~ . . . - - - ,_. . ,- . , - . - - .- -_ - - _ _ . _ _
BRP-CRDR-7 Page 3 of 3 Signatures The team member identifying the HED shall sign the completed form. In the absence of the original team member, the review team leader shall sign the form. After the team member signs, the form shall be submitted to the plant and signed as acknowledgement of an HED. Plant signature will be obtained after categorization. Closeout signature can be a team member if the correction is implemented before closeout of the project. Af ter closeout, a plant representative shall sign when the corrective action is complete. i I -1 i e J l t
i & ' e-
" BRP-CRDR-7B W
f) BIG ROCK POINT HED SEQ No. y HUMAN ENGINEERING DISCREPANCY (HED) { d EED CATEGORY: Discrepancy Information
- 1. HED Title From Review k
of f
// h '2 Cross referencing
, Item No.
f Panel ID j System / Subsystem Modality: O Vision O Noise O Distraction g C Readability O Audibility O Display Adequacy ,d 3. Description 'f l.
- 4. NUREG 0700 I.D.
- 5. Photo Log No ,
Photo Title / Code Classification Information:
- 1. Specific operator error (s) that could result from HED I
l l
- 2. Procedures or operations that use the listed item in a manner to produce error l
Doc ID 8715A 517A
- 3. Cons;qu nc o of operator error (all modes)
Corrective Actiont
- 1. Suggested Resolution
- 2. Schedule 1
General Comments: r-s j-I Team Rep. - Plant Rep. Closeout Doc ID 8715A S17A 4 gw---- , , n - a - , -, n-- - - - n ,_r - - - - - e e,, ,,w,,rww--- _, n -- - - - -- - -
BRP-CRDR-78 liiD SEQGiNCE NO: 0001 Revlon No: HED CA N : BIG ROW POItG lif1AN ENGIIE! RING 1ESCREPNCf OED) DISCPEPNJCY I! FORMATION
- 1) HED
Title:
Hand Switch anWaent function From Review: General Panel Checklist BRP-CRDR-8D1 Item G-6 Z) Cross Referencing: Item No: HS-7902 Sys/Ref Dwg: ECS G40107 Panel ID: C01-A Modality: Vision Noise Distraction x Readability Audibility Display Adequacy 3)
Description:
Red light not identified for valve position. No red / green lite convention for valve position and no permanent name-plate. Function emergency condenser makeup valve. I
- 4) ICREG 0700 ID: 6.6.3.1.A i 5) Photo Log No:
Photo Title / Code:RXC-1 II CLASSIFICATION ItFORMATION: l' ! 1) Specific operator error (s) that could result from IED: Failure to identify valve position N/A Z) Operations that use the listed item in a manner to produce error: N/A l
- 3) Consequences of operator error (all modes):
N/A l 1 !t ! CORRECTIVE ACTION:
- 1) Suggested Resolution:
l Permanent nameplate installed below red light, "Emerg Cond Makeup Valve SVO47 Energized." Z) ScheckJ1e: Complete GETERAL COMMDES: Plant Rep Team Rep Closecut: Date
BRP-CRDR-7B HED SEQ No. BIG ROCK POINT HUMAN ENGINEERING DISCREPANCY (HED) HED CATEGORY: Discrepancy Information:
- 1. HED Title 4
From Review
- l. 2. Cross referencing Item No.
I^
+
Panel ID System / Subsystem
- i. '
Modality: [:) Vision C:] Noise [ ) Distraction I [:3 Readability E~) Audibility C:) Display Adequacy
- 3. Description i
i 1 l
- 4. NUREG 0700 I.D.
ji 5. Photo Icg No Photo Title / Code
' classification Information:
- 1. Specific operator error (s) that could result from HED l
2. Operations that use the listed item in a manner to produce error I l Doc ID 8715A 517A
- 3. Consequences of operator error (all modes)
Corrective Action:
- 1. Suggested Resolution
- I I
- 2. Schedule i
a-General Comments: i I P I i 1 , l' l j Team Rep. - Plant Rep. i
- Closeout r
Date i Doc ID 8715A 517A t i
,,-w - - - , - - - . , ,,,..-----.n, ,, , - , . , .,,n.--nw-.-- - - - - , . - , , . - - , - - , - , _ - , , . _ , - . . . . - - - - - - - - _ - , - , - , - - , , - , - . . - - --
p ATTACHMENT K f HED Categorization Procedures t t
BRP-CRDR-7A Page 1 of 2 HED CATEGORIZATION HEDs identified during the.various surveys, reviews, task analysis and checklists will be evaluated by the review team collectively for their potential impact on emergency and normal operations. Four categories each designed to be unique will be utilized. A consensus shall be obtained from the team members. Highest to lowest priority shall be 1 to 4. UfRDRT w~ _ Criteria for each category Category 1 HEDs that are judged likely to adversely affect safety or the
};
management of emergency conditions by control room operators.
' HEDs must be of the DOCUMENTED nature (ie., LER's, historical data, interviews, questionnaires, etc).
Level A includes those HEDs for which the related documented error was associated with a safety important function, or,any function necessary to mitigate the consequences of an accident.- Also the HEDs must identify that the documented error DID result i in unsafe operation or the violation of a technical specification. Level B includes those HEDs for which the related documented error was associated with a safety important function, or any function necessary to mitigate the consequences of an accident. Also the HEDs must identify that the documented error DID NOT (but could have) result in unsafe operation or the , violation of a [; technical specification.
,. Category 2 HEDs that are judged likely to adversely affect safety or the i management of emergency conditions by control room operators.
HEDs are of the UNDOCUMENTED nature (surveys, etc). j ' Level A includes those HEDs for which the related undocumented
- l. error was associated with a safety important function or any function necessary to mitigate the consequences of an accident.
Also the HEDs must identify that the undocumented error COULD (but did not) result in unsafe operation or the violation of a technical specification. Note: Since the potential for error was undocumented there can not be levels identifying occurrence of the event either leading to or
' not leading to an error associated with a safety.important function. Consequently the way category II is defined, there can only be one level or scenario which is applicable.
Doc ID 8816A 4
l-BRP-CRDR-7 A Page 2 of 2 Category 3 HEDs that are not important to a safety function be it documented or undocumented. Level A includes those HEDs for which the related error was associated with a function not important to safety and DID (implies documented) result in an operational error which does not violate a technical specification or create an unsafe operating condition. j" Level B includes HEDs for which the related error was associated
' with a function not important to safety and COULD (implies undocumented) result in an operational error which would not 7 violate a technical sp.2cification or create an unsafe operating condition.
Level C includes HEDs for which the related error was associated with a function not important to safety and DID NOT or COULD NOT result in an operational error which would not violate a technical specification or create an unsafe operating condition. Category 4 Includes HEDs which are not important to a safety function that has been assessed and determined neither to increase any potential for causing or. contributing to operator errors.
'. Note 1: There are no levels associated with this category and this
. category differs from 3C in that this type HED does not increase the potential for error.
L'. Note 2: This category, along with the others, shall be analyzed for interaction of HEDs to determine conflicts, cumulative risk of error and compounding errors. J m
E
\
/
f / a 5 s ATTACHMENT L Control Room Design Conventions for Control's~ O t
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ATTACHMENT M ' l Control Room Design Modification Procedures and Forms i-9 4 j. t e a e
Ravioica laa Pag 2 19 cf NUCLEAR OPERATIONS DEPARTMENT PLANT OPERATION / MAINTENANCE f( h/NOA CONTROL OF MODIFICATIONS g4 g Jg jr TITLE: details. The Safety Evaluation shall identify whether or not the FC Minor constitutes an unreviewed safety The question, safetyper 10CFR evaluation 50.59 (changes, tests and experiments). responsible shall be reviewed by a competent person who was not for its preparation. The RE then shall sign the FC Minor, Form 3103, for completion of conceptual design. 5.4.4 0-List Determination The E&MM/S shall review the FC Minor package to determine if the modification involves or interfaces with Q-listed components, If the equipment or activities, in accordance with NODS-A21. modification is determined to be Q-listed, the FC Minor package shall be submitted to the Plant Review inor Committee package shall (PRC) for review If the modification is not Q-listed, the F proceed for detailed design, per Sectio {" g(, ( i 5.4.5 PRC Review
- The Q-Listed FC Minor package and its safety evaluation shall l receive a PRC review in accordance with NODS-A17. The RE include evidence of the PRC review within the package.
I 5.4.6 Detailed Design l I The RE shall proceed with the detailed design required to supportThe d the design input identified on the DIC. demonstrate adherence to referenced codes and standards, design bases and regulatory requirements as required, or demonstrate i eer-
- technical adequacy of the design with accepted industry enc n j
ing practices. Changes or exceptions toDetailed identified inputs, design documen-including reasons, shall be documented. tation shall consist of Engineering Analysis, Form 3119, dies, ble. equipment / material specifications, drawings, proce y monitor the detailed design process lfor The RE shall continuall' any tion. impact or change that would alter the approved s shall prepare a new safety evaluation and obtain the appropriate i ith approvals per NODS-A17. The RE should review the modificat on w Operations Department Operators to discuss the modificationinput as l features and functions, and to obtain Operators'
, practicality and preference which would accommodate equipment operation.
N00186-0005A-QA06
].3 y ;, gg j cq,e (
Std No NODS-P08 R2vicicn 13A NUCLEAR OPERATIONS DEPARTMENT Pcge 33 of PLANT OPERATION / MAINTENANCE TITLE: CONTROL OF MODIFICATIONS 5.5.7 Detailed Design - Output ([k.[fittb Ndhiwd) The Project Engineer shall review the modification with Operations l Department Operators to discuss the modification features and functions, and to obtain Operators input as to the practicality and preference that would accommodate equipment operation,
- a. The Project Team shall proceed with the detailed design required to support the design input identified on the Design Plan. The detailed design shall demonstrate adherence to l! referenced codes and standards, design baser and regulatory requirements as required, or demonstrate technical adequacy of the design with accepted industry engineering practices.
Changes or exceptions to identified inputs including reasons, shall be documented. Detailed design documentation shall consist of engineering analyses, equipment / material specifi-cations, drawings, procedures, studies, and other documenta-tion which supports the design, as applicable. Detailed Design output documents shall be prepared and re-viewed as follows: I 1. Engineering analyses may include physics, seismic, radia-tion, stress, thermal, hydraulic and accident analyses;
, assessment of material compatibility; and determination of accessibility for inservice inspection. Engineering analysis documentation shall include:
(a) A title which clearly identifies the structure, systems or components to which the analysis applies I' i' (b) Identification of reference documents, such as l drawings instructions or procedures , including revision status t (c) The objective of the analysis, if not clear from the title (d) The design input which is being considered (e) The results of literature searches or other applica-ble background data (f) Assumptions used
,N00186-0005A-QA06
. CORSumtf5 Power BIG ROCK POINT NUCLEAR PLANT F C-Company Oesign input checklist
- 1. DESIGN REFERENCE DOCUMENTS identify the documents or sections from which applicable design criteria / constraints are given. Either reference or include such in the desig') package.
Applicable Reference A. MlOSHA B. Technical Specifications C. Codes D. Standards E. Regulatory Requirements F. Regulatory Guides G. Plant Engineering Specifications H. FHSR I. O-List i J. Plant Drawings K.
- 2. DESIGN EXTERNAL-ENVIRONMENTAL CONDITIONS Identify the enternal conditions which will affect items, systems, structures or components in this design. Discuss the affect and design considerations for such conditions. Documented environmental qualification of harsh environment safety-related electrical equipment is required by Tech Spec Section 6.
Applicable Reference A. Pressure B. Temperature
, C. Chemical / Corrosion / Protective Coatings D. Humidity f
E. Radiation /Effect on Material / Personnel Exposure / Location or Rad Zone F. Flooding ___ i G. Harsh Environment 10CFR 50.49:IEEE 3231974 Pipi Whip Inside Containment Reg Guide 1,46 i H. f I. e
- 3. DESIGN STRUCTURAL LOADS f identify the structural forces expected to be satisfied by this design. Discuss considerations.
1 1 Applicable Reference A. Seismic B. Wind C. Thermal l D. Dynamic - E. Static F. Anchorage IE Bulletin 80-21 G. l l Form 3123 8 83
e
- 4. MECHANICAL REQUIREMENTS Identify the mechanical requirements of this design, including appropriate calculations as necessary and considerations of vendor specs versus existing specs.
Applicable Reference A. Vibration B. Stress C. Shock D. Pressure E. Temperature F. Reaction Forces G. Special Handling H. Material Compatibility I.
- 5. HYDRAULIC REQUIREMENTS
, Identify hydraulics and pump / flow /6 p requirements. Include calculations as necessary.
1 Applicable Reference A. NPSH B. Op C. Velocities D. Capacity
. E.
I I
- 6. INSTRUMENTATION ldentify instrumentation required considering installation, testing and operation of the system. l&C Engineer should be consulted.
Applicable Reference A. Power Supply Loading (DC. Air Supply)
- 8. Testability of Equipment (Standards)
C. Range' Span / Dead Band, etc D. Piping' Tubing Layout l'iE. Controls / Alarms / Set Points I ' F. Human Engineering -- I G. - - -
- 7. ELECTRICAL REQUIREMENTS
, identify electrical requirements. This section should include load studies, breaker timing studies. etc, as appropriate. Type of wire / current voltage considerations should be discussed.
Applicable Reference A. P:wer Supply Loading B. Voltage / Current C. Insulation / Wire / Cable D. Motors E. Raceway (Separability. Redundancy) 10CFR 50 Appendix R: lEEE 384 F. NEM A Class Requirements G. Containment Penetrations Reg Guide 1.63 l H. _ . _ _ . . _ _ _ _ _ _ _ _ . _ . . _ _ . . _ . _ _ _ . . i
- 8. F1,.E PRLTECTION ret UILEMENTS The requirements for material flammability, separation and controls (including installation) should be discussed. These consid-erations should conform to BTP APCSB 9.5-1,5/196, see NRC correspondence dated 9/2806.
Applicable Reference A. Materials B. Separability C. Fire Barriers D.
- 9. CHEMISTRY REQUIREMENTS Identify chemistry conditions which must be satisfied or specified in the new design. Chemical Engineer should be consulted.
Applicable Reference A. Corrosion B. Water Chemistry Limits __,
! C. Lay.Up D. Sampling _._
E. Hazardous Materials F. _,
- 10. SECURITY / ADMINISTRATIVE REQUIREMENTS
{ Any special controls such as work in vital areas, etc, may require special security requirements. Property Protection Supervisor
^
sh:uld be consulted. Applicable Reference
. A.
l B. _ _ _ _ . C.
- 11. PHYSICAL INTERFACES In performing the work associated with this project, there may exist interface requirements that physically impede work or result in special considerations including future maintenance. Potential interfaces and required controls should be considered here. Special consideration should be given to core drilling and digging.
Applicable, Reference A. Other Structures B. Other Components C. .
- 12. FUNCTION AL INTERF ACES The design addition or modification may functionally affect the operation of other systems / components. Additions of breakers
?- 13 a bus may functionally interface with the timing of other breakers. The addition of flow orifice, etc, may have other system functional interfaces. Address these interfaces if not covered in previous requirements. Consideration should be given to opera-tion in various plant modes.
Applicable Reference A. Other Systems ~ B. Other Components , , _ l C. Operationat Recuirements Under Various Plant Modes O- - - . . .--- -- _ . . . . _ - - . _ . ..
- 13. DISMANTLING AND REARRANGING OF EXISTING OR NEW ITEMS To perform this specific design, other existing plant items may have to be temporarily moved, dismantled or retired. If so, apptrpriate design information should be provided herein. These sketches, etc, address the information normally submitted with maintenance orders for in plant minor modifications.
Applicable Reference A. Sketches B. Instructions C. Procedures D. Special Handling E.
- 14. TESTING REQUIREMENTS With the design, if ASME Codes are involved, certain preservice and inservice tests are required. They should be addressed in this category. Acceptance testing requirements should be specified here as well as any new surveillance test requirements.
I, Applicable Reference A. Acceptance B. Surveillance C. Preservice D. Inservice E.
- 15. ADDITIONAL OA/OC REQUIREMENTS Normally, code, specifications, etc, produce the required QA/OC controls; however, for certain jobs or special processes not covered by the codes, additional OA/OC requirements may be desired. The case of a non-Q job with high reliability aspects f and a desire for OC is an example.
Applicable Reference B. C.
- 16. SCHEDULING REQUIREMENTS i.
Grod project engineering practices dictate scheduling considerations. In the case of outage related work, mini schedules should be included to address overall planning as a part of the project. Routine work to be performed by Plant Maintenance which does not affect outage activities does not require Project Engineer involvement. Applicable Reference A. B. C. Performed by Date Technical Review by Date
Ta Troisi:rg Department From E&M Superintendent / CONSUMERS POWER Date COMPANY Subject BIG ROCK POINT PLANT Internal NOTIFICATION OF MODIFICATION Correspondence FC/SC/SPC AND THE NEED FOR TRAINING CC Operations Supervisor FC/SC/SPC Package f Maintenance Supervisor ~ Chemistry / Health Physics Superintendent Shift Supervisors Office PRA Engineer In accordance with the Engineering Manual, Volume 16, Chapter 16.5, this is to inform you of the subject modification and its training requirement. Title of Modification: Training Required: Related Modifications (ie, FC/SC/SPC): _
References:
Attachments:
~
Planned Construction Start Date Finish Date l Description of Modification: 9 l k e 9 9 I l BP0784-0006A-BT01
UEE NUCLEAR OPERATIONS DEPARTMENT ' N Facility Change company u jor modification Plant UFlNo. , , , , , Plant System Code , , System Eau.oment
Title:
Processing Sequence Date Plant Project No. Oi ' A'd GWO/MWO No. j Functionat
Description:
]'Y N l
t Tech Suoti PRC Action Mtg No./Date
;oncept Design i
Pri Con tact
,DDC Complete Plant Engr CWPs identified I
~ ~ ~ ~ ~~
Plant Engr
^ L AR A Review Spare Parts Ordered Pri Contact Completed, inspected. Tested
, Plant Engr _.. . _ . . . . . _ . ._. _ _ _ . . _ - . _ . Required Docs issued
. . . . . . _ .__ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ . _ _ . _ _. Plant Engr QA NO DeficiencieI QA Suot Operability Authorization
. _ _ _ . . _ _ . _ _ ._ Y Ops Supt Punch List Dispositioned
- - _ _ . _ . - . .. __ ._..._ _ _ Tech Suot Closcout Tech Suot, Firm 3170 2 82
EE NUCLEAR OPERATIONS DEPARTMENT g Fac,lity Ch:nge
-[ ,
1 gg i Minor Modification. Page of Y: ject No. Plant i UFlNo. , , , l , , , , Plant System Code , , System Eauipment Processing Sequence Date _ Title : Prelim Enge Comp Functirnal
Description:
Proi Engr' Q-List Affected Y N Tech Suot PRC Action Mtg NoJDate Detailed Engr Comp Proi Engr Pros Review Comp i l Tem Suot ALARA Review l l 1 QA Review ; t QA Authorization To implement i I Plant Mgr/ Supt' Operations Notified 4 Proi Enge Completed, inspected, Tested l 1
--. Proi Engr!
QA Review l l l l QA Supve
. ___. _..__. . - . - - - _ . _ _ operacility Authorization Ops Supt M m .e m
.--h6_ m _=4h e N Document Update ubeg Nu_ap .=Wmim'rh6 +eW _ e
_ _ _ . _ . . _ _ , _ ____ _ _ _ . CIOleout , i i Teer Saet' Form Q10) 1-S7
e ATTACHMENT N List of Meeting Attendees U s
.s.
CRDR EXIT BRIEFING January 31 1986, 0800 Name -Affiliation Title Bob Alexander CPCo Technical Engineer Leo Beltracchi USNRC Senior H. F. Engineer J. S. DeBor SAIC/USNRC Scientist P. M. Donnelly CPCo Risk Assessment T. R. Fisher CPCo SR QA' Consultant D. L. Fugere CPCo (G.0.) Emergency Planning S. Guthrie USNRC Sr. Res. Insp, BRP Ron Hicks CPCo Engr. Field Org. D. P. Hoffman. CPCo (BRP) Plant Superintendent. Ray Ramirez USNRC Senior H. F. Engineer T. S. Rotella USNRC Project Manager
! R. E. Schrader CPCo (BRP) Elect /I&C Eng. Supv.
. . D. H. Schultz Comex/USNRC Operations Engineer W. J. Trubilowicz CPCo (BRP) Operations Supervisor R
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Enclosure 2 SAFETY PARAMETER DISPLAY SYSTEM (SPDS) NOTES FOR CONSUMERS POWER COMPANY'S BIG ROCK POINT NUCLEAR POWER PLANT BACKGROUND On June 20, 1984 Consumers Power Company submitted a Detailed Control Room Design Review (DCRDR) Program Plan for Big Rock Point Nuclear Power Plant (Reference 1). Section 9.2 of the licensee's Program Plan described the methodology for a Safety Parameter Display System (SPDS) Justification Investigation to be conducted as part of the DCRDR. In Section 9.2, the licensee stated that because of the small size of the Big Rock Point control ! room, it would be prudent to utilize the results of the DCRDR as justifica-tion for either installing or not installing an SPDS. See Attachment 1 for the SPDS justification. On November 7,1984, the NRC Staff Comments on Big Rock Point Detailed Control Room Design Review (DCRDR) Program Plan were forwarded to the licensee (Reference 2). Though specific SPDS concerns were not identified in this report, the NRC staff did express important concerns regarding the overall thoroughness of the approach documented in the DCRDR Program Plan. The need for a meeting to resolve Program Plan concerns was identified in thi s letter. , On January 22, 1985, the NRC staff forwarded Additional Staff Comments on Detailed Control Room Design Review (DCRDR) Program Plan to the licensee (Reference 3). This letter included the NRC staff's comments regarding the results of additional Program Plan review activity. The purpose of the additional NRC review was to assess the adequacy of the DCRDR function and task analysis for determining whether an SPDS is necessary at Big Rock Point. The additional review concluded that the scope of the DCRDR for evaluating the need for an SPDS may be inadequate. As a result, the NRC staff recommended that " Radiation Control" be included as a function within the function and task analysis activity and validation of control room functions. The NRC staff also recommended that the licensee study Supple-ment 1 to NUREG-0737 (Reference 4) in order to evaluate the licensee's SPDS needs in terms of the five required functions: 1
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(1) Reactivity control (ii) . Reactor core cooling and heat removal from the primary system (iii) Reactor coolant system integrity (iv) Radioactivity controls (v) Containment conditions In order to prepare for an NRC evaluation of the adequacy of the DCRDR to assess the parameter needs for an SPDS, NRC requested that SAIC review the Big Rock Point systems and parameters. Big Rock Point is a "one of a kind" 75 MW boiling water reactor. Operated since 1962, it is one of the least-automated reactors in the U.S. -utility network. The systems such as Steam Drum, Poison Tank, and Emergency Condensors are different than typical GE BWRs. BIG ROCK POINT SPDS NOTES The NRC DCRDR in-progress audit team planned to evaluate the licensee's SPDS justifications during the audit (January 27-31, 1986). However, as a result of lengthy discussions with the licensee during the audit, it became apparent that they misunderstood the SPDS requirement in Supplement 1 to NUREG-0737. The licensee was under the misconception that they were being required to install a new multimillion dollar computer system in order to satisfy the requirement. NUREG-0800 (Page 18.:2-A8) states that " functional criteria for the SPDS do not rule out the use of other types of displays in SPDS designs. Review guidelines for specific SPDS designs that do not use CRT displays will be developed case by case, as they are needed" (Reference 5). - The licensee made a verbal commitment to design an SPDS that meets the unique needs of Big Rock Point. But, in order to develop an SPDS, the licensee was advised that they need to prepare a safety analysis report for NRC. The documentation and NRC review requirements presented in Supplement 1 to NUREG-0737 are- as follows:
- a. The licensee shall prepare a written safety analysis describing the basis on which the selected parameters are sufficient to assess the safety status of each identified function for a wide 2
O range of events, which include symptoms of severe accidents. Such analysis, along with the specific implementation plan for SPDS shall be reviewed as described in NUREG-0800. s
- b. The licensee's proposed implementation of an SPDS system shall be reviewed in accordance with the licensee's technical specifica-tions to determine whether the changes involve an unreviewed safety question or change of technical specifications. If they do, they shall be processed in the normal fashion with prior NRC review. If the changes do not involve an unreviewed safety question or a change in the technical specifications, the licensee may implement such changes without prior approval by NRC or may request a pre-implementation review and approval. If the changes are to be implemented without prior NRC approval, the licensee's analysis shall be submitted to NRC promptly on completion of review by the licensee's offsite safety review committee. Based on the results of NRC review, the Director of IE or the Director of NRR may request or direct the licensee to cease implementation if a serious safety question is posed by the licensee's proposed system, or if the licensee's analysis is seriously inadequate.
These summary notes reflect the NRC's findings on the DCRDR-SPDS audit conducted at Big Rock Point.
- 1. The control room is small (see Figure 1).
- 2. At the time of the audit, the licensee had identified six ,
variables for SPDS. These were: (1) Containment Sump Water Level (2) Containment Temperature, (3) Containment Pressure, (4) Primary System Pressure, (5) Primary System level, and (6) Power Level .
- 3. One of the key concerns that the licensee will have to address is the question of radioactivity control. This is not an E0P entry symptom and is therefore not considered a critical safety function by the licensee. However, it is a Supplement 1 to NUREG-0737 requirement. -
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- 4. Another concern is the containment temperature chart recorder location. The recorder is located on the back of panel CO2. Its current location is classified as part of the control room. But, the SPDS requirements call for a display system that is located convenient to the control room operators. In the past, this has not included instruments on back panels. The licensee will either have to move the containment temperature chart recorder or provide
-a very substantial justification for its location.
- 5. The licensee's list of parameters is incomplete when compared with the criteria (Reference 6) identified by the former Procedures and Systems Review Branch (see Figure 2). For example, there is no containment isolation status or emergency condensor flow to steam drum (pressure and temperature on return leg are needed as an indirect measure of flow).
f 6. Validation of SPDS parameters prior to display meets " reliable" ! portion of the SPDS requirement. However, the licensee's control f room does not have to meet Reg. Guide 1.97, Post Accident Monitor-ing. Therefore, if the licensee goes with an analog hardwired SPDS, it may not be practical to have the licensee meet the ! validation requirement. This is because redundant instruments wil,1 not be available in all cases. With an analog hardwired SPDS, validation of data is done by an operator using redundant instrumentation.
- 7. At the audit, the licensee did express thoughts on highlighting SPDS instruments by color coding the background of the control board in the vicinity of the instrument. If this is done, then it is very important to evaluate the dispersion among instruments needed to evaluate each critical safety function. If several instruments are needed to evaluate a critical safety function, and these instruments are located equally between the control panels and the back panels (not within the field of view while at the control console), then human factors principles (the integration of related variables) have been violated.
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APPENDIX B NRC PROPOSED SPDS PARAMETERS (MINIMUM REQUIREMENTS) - 8,WR Safety Function Parameter Reactivity Control (a) APRM SRM Core Cooling and Heat Removal RPV Water Level , Drywell Temperature (b) Pressure Vessel Integrity RPV Pressure Radioactivity Control Main Stack or Offgas (Pretreatment) Monitor Containment Radiation Monitor Conta.inment Integrity Drywell Pressure Drywell Temperature Suppression Pool Temperature Suppression Pool Level (c) l Containment Isolation Valve ! . Status Hydrog'en/0xygen Concentration ! HOTES: (a) Intennediate Range Monitors (IRMs) are not considered essential due to their limited range of applicability (<1% power). In some . cases SRMs and APRMs will provide overlap to cover the intermediate range. (b) Drywell temperature is significant for core cooling because of its potential adverse effect on reactor vessel water level instrumentation (adverse environment compensation). Figure 2 - B L1J R Parameters (R e f'e r e r7 c e 6) > 6
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- 8. The decision to use/ delay / bypass the Rapid Depressurization System (RDS) is critical as it impacts cooling of the core. Parameters needed for the decision must be integrated and must be rapidly and reliably evaluated by the operat'or. Furthermore, this problem is complicated because personnel are allowed in the containment while the reactor is operating; they must be evacuated and the contain-ment isolated upon use of the RDS in order to prevent uncontrolled release to the atmosphere. It is recommended that the RDS and its operation be analyzed to identify parameters that should appear on a Safety Parameter Display System.
- 9. Trend information from a multipoint recorder may be difficult to identify, read, and use. A potential violation of rapid and reliable use of the data may exist.
- 10. The licensee should be informed that the use of a digital computer to implement the SPDS is acceptable. The computer may be a process computer and need not meet IE criteria. A small personal computer with a CRT display would suffice to meet the SPDS requirements. However, the cost to wire sensors to the computer is a major cost item, which is greater than the cost of the computer hardware and software. -
CONCLUSION While the above issues are necessary issues to review during an evaluation of the licensee SPDS, they do not constitute a complete set needed for the Supplement I to NUREG-0737 review. The licensee made a verbal commitment to develop an appropriate SPDS concept and provide the NRC with an SPDS Safety Analysis Report as required by Supplement 1 to NUREG-0737. . 7
REFERENCES
- 1. Letter to: Dennis M. Crutchfield, NRC, from: R.M. Krich, subject:
Big Rock Point Plant - Control Room Design Review (CRDR) Program Plan, June 20, 1984.
- 2. Letter to: David J. VandeWalle, Consumers Power Company, from: John A. Zwolinski, NRC, subject: Staff Comments on Big Rock Point Detailed Control Room Design Review (DCRDR) Program Plan, November 7,1984.
- 3. Letter to: David J. VandeWalle, Consumers Power Company, from: John A. Zwolinski, NRC, subject: Additional S'taff Comments on Detailed Control Room Design Review (DCRDR) Program Plan, January 22, 1985.
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- 4. Supplement 1 to NUREG-0737 - Requirements for Emergency Response Capability (Generic Letter No. 82-33), U.S. Nuclear Regulatory Commis-sion, December 17, 1982.
- 5. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 18.2, and Appendix A to SRP Section 18.2 U.S. Nuclear Regulatory Commission, November 1984.
- 6. Memorandum for F. Rosa (N.RC), M. Srinivasan (NRC) from: G. Mazetis (NRC), subject: SPDS Variable Selection Reviews, NRC, November 22, 1985.
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f ATTACHMENT 1 DCRDR PROGRAM PLAN SECTION 9.2 SPDS JUSTIFICATION INVESTIGATION L.
1 . i' C I 1
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9.2 SPDS JUSTIFICATION INVESTIGATION The integration of the SPDS into the CRDR for 3RP is quite unique. As indicated in a CP Co submittal to the NRC which provided our response to Supplement 1 of NUREG'0737*, it is our firm belief that with the small size of the BRP control room, utilizing the results of the CRDR as justification for either installing or not installing a SPDS is a more prudent course of action than is installing a gPDS l l before the review. In order to develop an adequate justification for the SPPS, the following procedure will be utilized:
- 1. During the questionaire portion of the operating experience review, the operators will be questioned as to the acceptability of existing displays of critical Plant variables.
- 2. During the interview portion of the operating experience review, the operators will be questioned as to the need for another display of critical Plant parameters in a central location.
*Ref. 9 9-2 REV 1, 6/08/84 mil 283-2686m103-142
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- 3. During the functions and task an.elysis portion of the review, accident scenarios will be, selected to ensnre that the operator is required to assess critical system functions for which a $PDS is intended to provide information at BRP. The functions are:
. reactivity control, reactor pressure centrol, reactor level control, containment pressure control contairment tec:perature s
control and containment livel centrol.
- 4. During the validation of control room functions, the following will occur:
- a. Various combinations of specific symptoms will be built into the walk-through scenario programs which will require that operators quickly evaluate Plant safety status by assessing the aforementioned critical system functions and as a result, perfore the required tasks to take the necessary response actions to mitigate unwanted conditions such as core degradation.
- b. During the walk-throughs, review tean members will monitor and document operator assessment and resulting actions and will note any specific problems which may hinder the operator in his efforts to perform an expeditious assessment of Plant status and to take prompt mitigating actions. Such problems will be classified as EDs.
9-3 EV 1, 6/08/84 mil 283-2686m103-142
- c. During the assessment phase, the review team will evaluate the results of the walk-throughs to determine whether or not the existing instrumentation design meets the intent of the SPDS as defined in Supplement 1 to NUREG 0737 (ie, the existing instrumentation represents a concise dispisy of critical Plant variables to the control room operators to enable thes to assess the status of the Plant and in assessing whether abnormal conditions warrant corrective actions to avoid, or sitigate the effects of, a degraded core). Should the walk-through results show that corrective actions are required, alternate proposals will be reviewed utilizing probabilistic risk assessment cost / benefit evaluation to determine the most cost-effective solution to remedy the probles.
l - i l ? l 9-4 ! REV 1, 6/08/84 l sil283-2686m103-142 _ _ _ _ _ _}}