ML20138N926

Audit Rept of Dcrdr for Arkansas Nuclear One,Unit 1, for Oct 1985

ML20138N926
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/31/1985
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML20136F969	List: ML20138N926
References
CON-NRC-03-82-096, CON-NRC-3-82-96, RTR-NUREG-0737, RTR-NUREG-737 NUDOCS 8511060161
Download: ML20138N926 (52)

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AUDIT REPORT 0F THE t

• DETAILED CONTROL ROOM DESIGN REVIEW FOR ARKANSAS NUCLEAR ONE, UNIT 1 October 1985 Prepared by:

Science Applications International Corporation 1710 Goodridge Drive McLean, Virginia 22102 l . ,

Under Contract to:

U.S. Nuclear Regulatory Comission Washington, D.C. 20555 -

4 Contract No. NRC-03-82-096

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. 1 FOREWORD This audit report was prepared by Science Applications International Corporation (SAIC) under contract NRC-03-82-096 Technical Assistance in support of NRC Licensing Actions: Program III. The assessment was i

performed in support of the Division of Human Factors Safety Human Factors EngineeringBranch(HFEB). HFEB previously evaluated Arkansas Power & Light l Detailed Control Room Company's (AP&L) generic Program Plan for conducting (ANO). Units 1 and 2.

Design Reviews (DCRDRs) at Arkansas Nuclear One

- Because the AP&L Program Plan provided insufficient details, the NRC staff met with AP&L on May 2,1984, where additional information was provided to describe AP&L's Program Plan. A summary of this meeting along with NRC comments was prepared and transmitted to the license on June 7,1984. AP&L l

submitted the DCRDR Final Summary Report for ANO, Unit 1 on August 14, 1985.

Based on a preliminary review of this Final Summary Report, and on the fact that both ANO Unit I and 2 are using the same DCROR process, the NRC decided .

to conduct an on-site, pre-implementation audit of the DCRDR for ANO-1 and .

an in-progress audit of ANO-2. This audit was conducted on September 16-20 1985. This evaluation of ANO-1 is based upon both the Program Plan and Summary Report submitted by AP&L and the information provided b) the licensee during the September 1985 pre-implementation audit.

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I TABLE OF CONTENTS Section Page 1

BACKGROUND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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PLANNING PHASE . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1. Preparation and Submission of a Program Plan . . . . . . . . 3

2. Establishment of a Qualified Multidisciplinary Review Team . . . . . . . . . . . . . . . . . . . . . . . . . 3

, REVIEW PHASE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

1. Review of Operating Experience . . . . . . . . . . . . . . . 6

2. System Function and Task Analysis . . . . . . . . . . . . . . 8

3. Control Room Inventory . . . . . . . . . . . . . . . . . . . 11

4. Control Room Survey . . . . . . . . . . . . . . . . . . . . . 12

5. Validation of Control Room Functions . . . . . . . . . . . . 15 ASSESSMENT AND IMPLEMENTATION . . . . . . . . . . . . . . . . . . . . 16

1. HED Assessment Methodology . . . . . . . . . . . . . . . . . 17

2. Selection of Design Improvements . . . . . . . . . . . . . . 18

3. Verification That Selected Design Improvements Will Provide the Necessary Correction and Can Be Introduced in the Control Room Without Creating Any Unacceptable Human Engineering Discrepancies . . . . . . . . . . . . . . . 20

4. Coordinat' ion of Control Room Improvements With Changes Resulting From Other Improvement Programs . . . . . . . . . . 21 DESCRIPTION OF. PROPOSED DESIGN CHANGES AND JUSTIFICATION AND HEDs WITH SAFETY SIGNIFICANCE TO BE LEFT UNCORRECTED OR PARTIALLY CORRECTED .............................. 21

1. Proposed Schedules for Implementing HED Corrections . . . . . 22

2. Proposed Corrective Actions and Justifications for

's- HEDs to be Left Uncorrected . . . . . . . . . . . . . . . . . 22 CONCLUSIONS AND RECOMMENDATIONS . . . . . . . . . . . . . . . . . . . 2,4 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 APPENDIX A - HEDs LISTED IN VCs NE 2 0F THE

SUMMARY

REPORT FOR WHICH CORRECTItt Ji *IOK3 OR JUSTIFICATIONS FOR NOT 31 CORRECTING d'N T/a ySED . . . . . . . . . . . . . . . .

APPENDIX B - ANO PRE-IMPLEMENTATION AUDIT MEETINGS . . . . . . . . . . 36 ,

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e TABLE OF CONTENTS (Continued)

Section Page

• ' I APPENDIX C - DIFFERENCES BETWEEN NUREG-0700 AND AP&L CHECKLIST PRESENTED AT ANO-1 PRE-!MPLEMENTATION AUDIT ON SEPTEMBER 18, 1985 ................... 40 I

APPENDIX D - A BRIEF DESCRIPTIUON OF REMOTE SHUTDOWN CAPABILITY AND COMMENTS ...................... 46

.. APPENDIX E - TASK ANALYSIS INSTRUMENTATION REQUIREMENT FORM . . . . . 48 e

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AUDIT REPORT OF THE DETAILED CONTROL ROOM DESIGN REVIEW FOR ,

ARKANSAS NUCLEAR ONE, UNIT 1 This report documents the findings from a pre-implementation audit. conducted on September 16-20, 1985, and the evaluation of the DCRDR ,

Final Summary Report submitted to the Nuclear Regulatory Commission (NRC) on August 14, 1985, by Arkansas Power & Light Company (AP&L) for the Arkansas Nuclear One, Unit 1 ( ANO-1) Detailed Control Room Design Review (DCRDR)

'(Reference 1). The DCRDR review by AP&L at ANO-1 was conducted in accordance with a generic Program Plan submitted to the NRC on November 25, 1983, for performing DCRDRs for both ANO Units 1 and 2 (Reference 2). The NRC staff reviewed the Program Plan and forwarded their comments to AP&L on February 2,1984 (Reference 3). The AP&L Program Plan had insufficient

- details addressing the processes to accomplish the DCRDR objectives; there-fore, the NRC staff met with AP&L on May 2,1984, in order to obtain ad-ditional information to supplement AP&L's Program Plan. AP&L submitted the DCRDR Final Summary Report for ANO-1 on August 14, 1985. The decision was made by the NRC to conduct a pre-implementation audit of the DCRDR for ANO-1 in order to clarify certain aspects of the generic review process, and to confirm that the review had been conducted appropriately.

BACKGROUND E

Licensees and applicants for operating licenses are required by the Nuclear Regulatory Commission to conduct a Detailed Control Room Design Review (DCRDR). The objective is to "... improve the ability of nuclear power plant control room operators to prevent accidents or cope with acci-dents if they occur by improving the information provided to them" (NUREG-0660. Item I.D.) (Reference 4). The need to conduct a DCRDR was confirmed in NUREG-0737 (Reference 5) and in Supplement 1 to NUREG-0737 (Reference 6).

DCRDR requirements in Supplement I to NUREG-0737 replaced those in earlier documents. Supplement I to NUREG-0737 requires each applicant or licensee to conduct its DCRDR on a schedule negotiated with the NRC. Guidelines for 1

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conducting a DCRDR are provided in NUREG-0700 (Reference 7) while criteria for NRC evaluation of a DCRDR are contained in NUREG-0800 (Reference 8).

A DCRDR is to be conducted according to the licensee's own Program Plan (which must be submitted to the NRC). According to NUREG-0700, t'he DCRDR should include four phases: (1) planning. (2) review. (3) assessment, and (4) reporting. The product of the last phase is a Summary Report which must

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include an outline of proposed control room changes. their proposed sched-ules for implementation, and summary justification for human engineering discrepancies (HEDs) with. safety significance to be left uncorrected or partially corrected. Upon receipt of the licensee's Summary Report and prior to implementation of proposed changes, the NRC must prepare a Safety Evaluation Report (SER) indicating the acceptability of the overall DCRDR effort (not just the Summary Report). The NRC's evaluation encompasses all documentation as,well as briefings, discussions, and audits if any were ,

conducted.

The purpose of SAIC's evaluation is to assist the NRC in the technical assessment process by providing an independe'nt appraisal of the Arkansas Nuclear One. Unit 1 DCRDR process and results.

The DCRDR requirements as stated in Supplement I to NUREG-0737 can be summarized in terms of nine specific issues, a list of which provides a convenient outline of the areas covered in this evaluation. The nine issues are:

1. Establishment of a qualified multidisciplinary review team.

2. Use of function and task analyses to identify control room operator tasks and information and control requirements during emergency operations.

'3. A comparison of display and control requirements with a control room inventory..

4. A control room survey to identify deviattoi s from accepted human factors principles.

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5. Assessment of human engineering discrepancies (HEDs) to determine which HEDs are significant and should be corrected.

6. Selection of design improvements that will correct those discrepan-cies. -

7. Verification that selected design improvements will p/ ovide the )

necessary correction.

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8. Verification that improvements can be introduced in the control l room without creating any unacceptable human engineering discrepan-cies.

9. Coordination of control room improvements with changes resulting from other improvement programs such as SPDS, operator training. -

new instrumentation, Reg. Guide 1.97, and upgraded emergency operating procedures.

PLANNING PHASE .

l 1. Preparation and Submission of a Program Plan The NRC staff reviewed Arkansas Power and Light Company's generic i Detailed Control Room Design Review Project Program Plan submitted for both ANO Units 1 and 2. The NRC reviewed the Program Plan with reference to the requirements of. Supplement 1 to NUREG-0737 and transmitted comments to AP&L by memo dated February 2,1984. Since AP&L's generic Program Plan provided insufficient details, the NRC staff met with AP&L on May 2,1984, where additional information on the DCRDR process was provided to supplement the Program Plan. -

2. Establishment of a Qualified Multidisciplinary Review Team Both the Program Plan and the Summary Report for ANO-1 included a l description of the staffing and management that were established to conduct i the DCRDR Team. The structure and management of the DCRDR team appeared to

be flexible enough to permit a multidisciplinary effort. Management and

! administration of the AP&L DCRDR at ANO-1 were the responsibility of the l

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NUREG-0737, Supplement 1 Steering Committee. The Steering Committee is composed of five members represen.ing the ANO Operations Manager and Units 1 and 2 Operations Superintendents along with the AP&L Engineering Services General M: nager and the Project Manager for ANO. Its responsibility is to provide management assistance to ensure integration of major project ob-jectives in a cost-effective manner. Working in conjunction with the 4

Steering Committee was the NUREG-0737 Supplement 1 Program Coordinator who was responsible for planning and coordinating activities to ensure the l integration of all objectives to meet NUREG-0737. Supplement 1. The DCRDR '

I Team Leader reports to the Steering Committee through the -NUREG-0737

, . Supplement I coordinator. -

The ANO-1 DCRDR team consisted of a group of professionals from AP&L and Advanced Resource Development Corporation (ARD) with a wide range of skills necessary for the performance of the DCRDR. Based upon the resumes provided in Appendix C of the Summary Report, the members of the DCRDR team appeared qualified to perform DCRDR activities. Expertise of the team included: ,

e I&C engineering

e Nuclear systems engineering

! e Human factors engineering e Operations e Electrical engineering e Operations assessment.

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During the ' audit, a matrix was presented that indicated which team members were assigned to each task of the DCRDR program. From this matrix,

. it appeared that the appropriate skills were available and properly utilized.

Prior to beginning the review, team members were familiarized with the

general design and operation of the plant, relevant NRC documentation, and general human factors engineering principles and methodology through semi-i nars given to AF&L employees by ARD and through informal instruction. DCRDR team members also were encouraged to document dissenting opinions and were provided access to plant facilities, personnel, documentation, and other information as needed to perform their tasks.

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Additional support was provided to the DCRDR program from AP&L support organizations. During the assessment process, assistance was provided by the technical analysis staff, the plant computer support group and. the plant training department. The plant drafting department helped with the control room panel drawing updating process while the Operations staff, assisted during the validation and task analysis efforts.

The continuity during the review process appeared to be excellent with only one member of the original team having left since the review started.

However, his function was immediately assumed by his supervisor for the

. remainder of the review process. In summary, we believe that AP&L has satisfactorily met the requirements of establishing and supporting the

. . efforts of a qualified multidisciplinary DCRDR team.

REVIEW PHASE AP&L review phase plans and activities included:

1. Review of historical events and of operating experience -

2. Task analysis

3. Control room inventory

4. Verification of task performance 1 5. Validation of control room functions

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i r. 6. Control room survey.

The above activities are those recommended by NUREG-0700 guidelines as i contributing to the review phase objectives. Activities 2, 3, 4, and 6 i

contribute to the accomplishment of specific DCRDR requirements contained in Supplement 1 to NUREG-0737. Activities 1 and 5 are recommended by NUREG-i 0700 guidelines.

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1. Review of Operating Experience AP&L conducted a two-part operating experience review at ANO-1 to identify conditions which impact on the probability for those operator errors which could affect safe operation of the plant. The first part of the effort, the historical event review, included a review of the documented operating history of the ANO-1 plant to identify any recurring problems and an examination of industry-wide sources applicable to the plant. The second part of the review, the operating experience review, included the conduct of an operator survey and interviews with operators to obtain feedback on previous operating experience.

To accomplish the review of plant operating history and industry-wide experience, five sources of historical information were collected and reviewed. Industry-wide reports that included Licensee Event Reports (LER),

the Significant Event Reports (SER), and the Significant Operating Event Reports (SOER) for the past five years were reviewed by a Human Factors '

- Specialist (HFS). The in-house documentation included Unit Transient Reports and Transient Assessment Program (TAP) Reports. Because ANO-1 did -

not include a large number of events found in the industry-wide reports, the scope of the review was expanded to cover TAPS from all Babcock and Wilcox (B&W) plants. All available ANO-1 Unit Transient and B&W TAP reports were reviewed for the DCRDR, resulting in a total of 527 historical reports.

Criteria were developed to determine if the historical reports described and documented a problem specifically related to control room equipment, procedures, or personnel error. All reports that met any one of the criteria cited in the Summary Report were retained in the Historical Review Notebook that documented ANO-l's historical experience review. These, reports were then recorded onto a two-pag'e Problem Analysis Report (PAR) that captured the pertinent information necessary to analyze each event and determine if the problem was applicable to ANO-1 and uncorrected at the plant. Using this process, the historical event review resulted in the surfacing of many problem areas; however, the majority of these problems were not applicable to ANO-1 or were previously addressed and resolved by AP&L. As a result, only two problems were evaluated as uncorrected at ANO-1 and were documented as HEDs.

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The ANO-1 operator survey effort entailed the administration of an open-ended, self-administered questionnaire to 48 staff members, including l all licensed operating personnel, operators in training class for ANO-1 and '

instructors licensed on Unit 1. The survey was structured to add'ress the nine content areas suggested in NUREG-0700, Section 6. The objective of the survey was to obtain pertinent knowledge that operating personnele'at ANO-1 possess regarding both positive and negative control room features which they had experienced and/or, observed in the . course of preparing for ,

operations or during operations. ,

Forty-eight surveys were distributed for completion, twenty-five (521) were returned by mail to the HFS. Confidentiality was ensured by assigning i each questionnaire a number rather than a name. The list of potential respondents and corresponding numbers were kept in confidence by HFS.

i As some information relevant to operator experience could not .be solicited easily by using a structured questionnaire format, subsed uent i

! - individual semistructured interviews were conducted by the HFS with 18 of the questionnaire respondents and two operators who had not previously

! responded to the questionnaire. The objectives of the follow-up interviews

! were (1) to clarify ambiguities in an individual's written responses to the self-administered questionnaire; (2) to gather additional details pertaining to that individual's responses; and (3) to address issues on which the 4 questionnaire responses revealed particular concern or for which no consen-sus emerged. .

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operator survey activities were consolidated and summarized using descrip-i tive statistics. Component and system information resulting from the

! interviews was used to enhance the questionnaire data where appropriate, and

the comments were organized by topical areas to facilitate the review by the DCRDR team at a later date. Findings from the operator survey resulted in i

the identification of 103 HEDs of which 23 were rated as safety significant (Category 1).

In summary, AP&L's operating experience review at ANO-1 appeared exten-  ;

f sive, thorough, and appropriately conducted. Consistent with NUREG-0700 ,

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objectives and guidelines. it entailed a systematic examination of industry-wide reports and plant-specific documents. Structured questionnaires and semistructured interviews were administered to and conducted with a satis-factory range of operating personnel. The activities conducted resulted in the identification of 105 HEDs which were not identified as a result of 1

other DCRD81 activities.

2. System Function and Task Analysis i

The objective of the ANO-1 system function and task analysis was to establish the input and output requirements of control room operators' tasks performed under emergency conditions. The basis for system function

analysis for ANO-1 was derived from the Abnormal Transient Operating Guidelines (ATOG), generic material from B&W. This ATOG document includes the following studies

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e Safety Sequence Diagrams (SSDs) developed to identify the specific 1 plant systems and subsystems including necessary operator actions used to accomplish safety functions.

4 e Event Tree Diagrams constructed to determine the various plant ,,

j conditions which could evolve following a postulated initiating event.

e System Auxiliary Diagrams (SAD) developed to identify the support-l j ing equ,ipment which must operate in order to support the functions j of the front-line systems and subsystems.

I i The Safety Sequence Diagrams presented the specific plant actions that could occur in order to satisfy the accomplishment of a safety function.

i The event tree diagrams were developed for selected basic types of tran-sients to account for all of the combinations in which plant actions might occur. The System Auxiliary Diagrams describe the components and supporting systems and subsystems which are necessary to enable each major system to

perform properly. Using the Transient Operating Guidelines, the ANO-1 i plant-specific E0Ps were then developed. Subsequently, these plant-specific j E0Ps were used as a starting point to identify tasks to be conducted by operators during emergency situations. To accomplish the objectives of the j

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DCRDR task analysis, a detailed breakdown of the operator functions into tasks was performed at each stage of the symptom-oriented procedures. A task description breakdown was completed for each tabbed section of the ANO- '

1 E0Ps for the four sets of observable symptoms listed below: l

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1. Loss of Subcooling Margin l

2. Inadequate Primary to Secondary Heat Transfer (overheating)

3. Excessive Primary to Secondary Heat Transfer (overcooling)

! 4. Inadequate Core Cooling.

In addition, certain additional procedures were used to generate additional tasks and subtasks that might be performed in the control room

'during abnormal conditions. These procedures include the following:

I Unit 1 Procedures Diesel Generator Operation Reactor Building Ventilation i Plant Startup Fire Protection System Remote Shutdown Reactor Coolant Pump and Motor Problems l Plant Shutdown and Cooldown Loss "of Service Water

Loss of DHR Load Reject '

Turbine Trip Loss of P.eactor Coolant M/U i

Pressurizer System Failure Loss of S/G Feed l N/C Cooldown Loss of Condenser Vacuum l CRD Malfunction I .

l Once the site-specific documentation was developed, operator actions which were implied or stated were written as task statements on the Task i Description Form. All unique tasks were identified, coded with a task

number, and grouped into the prevailing system being exercised or ac,ted upon.

The tasks were broken down into task elements and/or aciiton steps by

! AP&L Operations staff subject matter experts (SMEs) in order to reflect a j step-by-step procedural set of actions that must be carried out in order to accomplish the task. These task reduction activities were accomplished by an iterative process involving a series of questions about each task, e.g.,

task conditions, initiating cues, frequency, and performance criteria. The

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t information and control needs for task performance were first collected on Task Description Forms and later entered into a database. This process resulted in the primary database for the entire DCRDR. l i '

j The above process was performed outside the control room as much as l possible. The task analysis performed for the DCRDR was not done from a l "what exists" perspective but rather in terms of "what should,be." The j process was not accomplished completely independent of the control room as i the SMEs also utilized procedures, piping and instrumentation diagrams, and

• electrical schematics as reference sources. However, the DCRDR team member's

,, pursued an iterative process and continued to probe SMEs for responses that  ;

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reflected "what should be" in the control room.

Task Analysis Instrumentation Requirement Forms were used to document

the information and control requirements generated from the selected pro- -

f cedures. This was done by an HFS and an Operations SME outside the control room. Like the task elements, the information and control requirements were coded from a "what is needed" perspective for the various action steps' j.

identified in the task analysis. The level of detail recorded for instru-mentation required, as shown by the " Task Analysis Instrumentation Require-

ment Form" in Appendix E, was found to be satisfactory. During the verifi-

! cation process, the availability of information and control requirements j were checked with existing instrumentation (inventory) in the control room. [

f , If the required instrumentation was found to be absent, then a " dummy" code j number representing that particular item was entered into the database and l resulted in the generation of an HED. Similarily, during the suitability comparison, if the required characteristics of displays and controls did not j match what was physically available in the control room, another HED was i i written. Forty-three HEDs were identified as a result of this process.

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i In conclusion, the AN0 Unit I system function and task analysis was l conducted in a comprehensive and systematic manner. The analysis was based on AN0's symtomatic Abnormal Transient Operating Guidelines (ATOG) Program J resulting in plant-specific E0Ps that were further defined into emergency j task elements and/or action steps. The task analysis covered not only the emergency operations but also selected abnormal and normal operational I

procedures such as Reactor Startup, Reactor Shutdown, and Cooldown resulting in broad coverage of systems necessary to support emergency operation.

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Subsequently, information and control requirements and their associated characteristics were derived for each action step. During the verification phase, availability and suitability comparisons were made with existing control room components, and appropriate HEDs were generated. This process resulted in an integrated task analysis that meets the requirements of Supplement I to NUREG-0737.

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3. Control Room Inventory The ANO-1,jnventory process used by the licensee included all displays, controls, indicators, and annunciators in the control room primary operating area, and was followed by a verification of availability and suitability as required by Supplement I to NUREG-0737.

4 The inventory was completed primarily by an HFS with assistance from AP&L 6perations staff as required in a process consistent with NUREG-0700 guidelines. The objective of the ANO-1 control room inventory was to establish a reference set of data which identified all instrumentation,

. controls, and equipment within the control room for comparison with the display and control requirements identified during the task analysis.

All displays, controls, controllers, annunciators, and other equipment in the control room with which the operators interact, were included in the com'prehensive inventory. A current set of as-built drawings were marked up to reflect the present configuration of the ANO-1 control room. A system-atic inventory was developed on a panel-by-panel basis where instruments were assigned a coded sequence number, inventorted, and checked off the as-built drawings. Each piece of equipment on the control boards and its characteristics were identified by codes that had been used to characterize

. information and control requirements identified in the task analysis. The

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inventory data was stored in the computerized database management system to be used as the basis of the verification process. ,

The objective of the verification process was to ensure that operator tasks derived from the plant-specific E0Ps could be performed in the exist-ing control room with minimum potential for human error. There are two aspects to the verification process. First, as described in the system function and task analysis, it was determined whether appropriate equipment 11

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was available in the control room to perform each task required by emergency operations. A total of 29 HEDs were identified during the verification of availability. Second, for equipment that had been identified as available, a suitability determination was made by comparing the characterist.ics iden-tified during the task analysis phase and the control room instrumentation documented during the inventory. This step was performed using the database by " matching" the two lists. Any "no match" items were noted as deviations and an effort was made to resolve these discrepancies. Those deviations that could not be resolved were recorded as HEDs resulting from a lack of control room item suitability. A total of 14 HEDs were recorded as a result of the verification of suitability.

In conclusion, AP&L has provided a detailed description of what appears to be a well-planned and well-executed control room inventory for ANO-1 that

. helped to produce a total of 43 HEDs from the verification phase of the DCRDR. The documentation process of the control room inventory utilizing the database management system was effective, and well-integrated, and produced a successful control room inventory that meets the requirements of Supplement 1 to NUREG-0737. . . ,

4. Control Room Survey A comparison of instrument and control features to the ANO-1 human factors guidelines was conducted. These guidelines were derived from those gi.ven in Section 6 of NUREG-0700 and closely follow them in format and content. The ANO guidelines do differ from those in NUREG-0700 in that some of the items were quantified, or reworded, to make them clearer and more concise for evaluation. These modifications to NUREG-0700 guidelines, as shown in Appendix of this report, were reviewed in detail by the audit team.

The results are presented in the following section: ,

Review of Differences Between NUREG-0700 Guidelines and AP&L Checklist i Several differences between the survey guidelines of NUREG-0700 l (Section 6) and the AP&L survey checklist are based on (1) misprints in l NUREG-0700; (2) qualitative guidelines which AP&L chose to make quantitative  !

to improve objectivity; and (3) very minor differences on guidelines of lesser significance. We find these modifications to be acceptable.

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However, the audit team disagrees with the modifications made by AP&L on the l following NUREG .0700 guidelines as discussed below:  !

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1.2.2.d(2), Exhibit 6.1-6,1.2.3.b.1.2.3.d(2) .

i TEAM POSITION:

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' Extended functional reach is measured from a wall to the tip of the right index finger with.the arm extended and the right shoulder extended out from the wall as far as possible with the i left shoulder against the wall. In order to minimize the poten-tial for inadvertant activation of controls, the operator should l I not be forced to lean over the benchboard to operate controls on

the back portion. Since the measurement referred to in the guide-lines is taken from the front edge of the benchboard, it is not .

equivalent to the extended functional reach measurement. In fact, it is 8 to 10 inches less than an extended functional reach. . At 25 inches for control board depth, the guideline of NUREG-0700 has already accounted for some amount of bending by most operators.

1 1.2.3. f(2) j!

TEAM POSITION:

As in previous guidelines, if a measurement could be used, the

) guidel.ines would probably be'.that controls should not be farther than 25 inches from the front edge of the console. The reason a I measurement was not used is that, depending on the task difficulty 1

and duration 25 inches may be too great a distance to reach.

3.2.1.c I

i TEAM POSITION:

D If a quantitative value for auditory warning signals were to be

specified, it should be specified as some maximum value over ambient noise level, not an absolute value of 90 db(A). Depending i on the ambient noise level, 90 db( A) may,very well startle or 13 .

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cause irritation to the operator. In addition, intensity is not the only signal quality that might startle or cause irritation.

5.1.6.c(2) ,

TEAM POSITION:

t The meanings recommended in the guidelines of NUREG-0700 for the use of red, green, and yellow as codes were determined to be within the stereotypical expectancy of our society's population.

Any other use of these colors, as codes, should be justified on an operational basis taking into account the behavioral attributes of the current expected operator population.

6.5.1.g TEAM POSITION:

Tag outs in the for's of plastic covers can physically prevent actuation of a control while providing sufficient writing surface and format to inscribe all required administrative information? -

Military systems use what is termed as " safety pins" with streamers to physically prevent control actuation.

9.2.2.e TEAM POSITION:

Guidelines on control / display packages should not be deleted since typical control rooms (including ANO) usually incorporate several modular vendor panels (e.g., turbine control, rad monitor).

The audit also disclosed that a number of checklist items had not been completed and that there were some apparent discrepancies in connection with the color-coding evaluation procedures.

14

1 I Remote Shutdown Panel Survey The NRC' has strongly recommended that a human engineering evaluation of the remote shutdown capability be included within the scope of the DCRDR.

although not explicitly identified as a requirement in Supplem'ent I to i NUREG-0737. Therefore. ' members of the NRC audit team did review the ANO-1 I remote shutdown capability and felt that it has many problems doe to pro-l cedural, labeling, and lighting deficiencies. Appendix 0 provides a more detailed description of the remote shutdown capability and its perceived i f problems.

i l In summary, the survey effort was complete, covering the nine content areas suggested in NUREG-0700 (e.g., workspace, panel design, annunicator 4 warning system, etc.). Environmental conditions, including sound lighting, i and the HVAC system, also were surveyed and audited by the NRC. While -

i primarily using the guidelines in NUREG-0700 as the basis for their survey.

l AP&L did modify six guidelines to which the audit team does not agree. The ,

I control room should be rechecked for these six items and responses provided if it is not in compliance. In addition, all sections of the checklist.

2 including color-coding, should be completed properly. Although recommended for review in NUREG-0700, the remote shutdown capability, which was found to

contain many problems, was not considered as part of the AP&L DCRDR at ANO-l 1. Although the survey was thorough, there are some items that have not been l

completed, along with six guidelines that should be reevaluated in the control room before this effort meets the requirements of Supplement 1 to l NUREG-0737.

l 5. Validation of Control Room Functions AP&L Company conducted a validation review at ANO-1 to determine -

! whether the functions allocated to the control room operating crew could be

! accomplished effectively within both the structure of the established emergency procedures and the designing of the control room as it now exists.

l The validation process was performed by using walk-through techniques i on the ANO-1 simulator. The events which were used in the validation were:

a 1

15 i

t

- --,- - ,...-.. --- - y6---. -%-.r.m--,--rw, , - , , , .,,.y. .- -,w-_-,,.-- --,,,_r-----oi.,-,,cww--.--

i e Automatic reactor trip with no abnormalities  ;

e Rea.ctor trip with overcooling margin e Reactor trip with low subcooling margin e Reactor trip with overheating ,

e Steam generator tube rupture e Steam line break. ..

The validation process was arranged and supervised by an HFS with the assistance of the DCRDR SME. Operating crews consisted of ,two ANO operators. During the simulations, which were recorded on videotape.

- operators were instructed to call out ' relevant actions, directions of move-ment, the displays, and indicators used, as well as their responses. The events and recording efforts were terminated when the SME determined that the crew had successfully investigated the event.

After recording the events, an HFS and an SME jointly reviewed and analyzed the data on a step-by-step basis. During this review process , ANO-1 procedures were referenced, the HFS would stop the tapes for viewing as needed, and the SME would clarify operator actions and identify procedural steps. Where the HFS observed instances in which equipment availability, suitability or location could be enhanced, or in which operator uncertainty existed due to procedural ambiguity. HEDs were written to improve operator actions and provide for corrective actions when operator deviations were safety significant.

1 In conclusion, while using a real-time simulator, the licensee imple-mented a validation approach consistent with the guidelines of NUREG-0700.

The events chosen were consistent with those suggested in NUREG-0700 and exercised all control room workstations. From the analysis performed, six HEDs were generated.

ASSESSMENT AND IMPLEMENTATION HED assessment and implementation procedures are described in Sections 7 and 8 of the Summary Report. Volume 2. Sections 1-12 of the Summary Report present review findings (HEDs).

16 4

, .-. - , - - . - . - . , -. _ - - - . , - _ , . . ~ - _ _ _ - - - - - - . - - _ . . - . . . . , . , . . - . _ - -

1. HED Assessment Methodology The DCRDR Assessment Team reviewed and assessed each HED based on its impact on plant safety and operability. The report states, "This review included a formal assessment cf each HED and evaluation of the most appro-priate action to initiate in order to further pursue correction of the HED."

4 T

Each member of the Assessment Team was provided with a notebook I containing a complete set of HEDs for his individual assessment. Each member was also provided with HED Assessment Rating Forms containing 20 questions grouped under the following factors affecting plant safety, plant

! operability, and personnel safety:

i e Impact on physical performance e Impact on sensory / perceptual performance

! e Impact on cognitive performance o Interaction with task variables e Impact or potential impact on operating crew error e Impact or potential impact on plant sa 'aty. ,

Upon completing the review of each HED against the six factors.*the l

j Assessment Team member determined the safety significance of that HED based on the following categories:

I. HIGHEST SIGNIFICANCE: Could affect or has substantially affected i

a safety system or operator response during an emergency situa-tion.

II. 51GNIFICANT: Could substantially affect or has substantially i.-. affected a nonsafety system or operator response during routine, nonemergency operation.

a III. LEAST SIGNIFICANT: Could or has affected operator response in a nonsubstantial way.

! The HED Assessment Rating Forms also provided for evaluation of the cumulative impact of Category 3 HEDs to ensure that the level of signifi- l cance was fully considered.

l ,

i 17 1

6^

Upon completion of the individual team member's assessments, a compila-tion was made and a copy provided each team member. The Assessment Team then met to discuss the ratings they assigned individually to each HED and to reach a team consensus regarding final HED categorization.

2 The only concern on the part of the reviewers of the Summary Report was that attempts at HED solution were being made during the assessment, an activity not properly part of the assessment phase (Reference 1, page 7-1).

1 During the audit, this concern was disspelled by an AP&L explanation of the assessment process and by the absence of any documentary evidence that HED

' correction was considered in connection with the assessment process.

During SAIC's evaluation of the Summary Report, all HEDs were reviewed for appropriateness of categorization. All HEDs in Categories 1 and 2 and a sample of the HEDs in Category 3 were reviewed by the NRC audit team to validate the licensee's assessment processi. The results indicated that the assessment process was generally detailed and correct.

.The assessment process as applied was satisfactory and meets the

requirements of Supplement I to NUREG-0737.

2. Selection of Design Improvements

. The purpose of selecting design improvements is to determine correc-tions to HEDs identified from the review phase of the DCRDR. Selection of design improvements should include a systematic process for development and comparison of alternative means of resolving HEDs. Furthermore, according to NUREG-0737 Supplement 1, the licensee should document all of the

~

proposed control room changes. ,

Although Ap&L's Summary Report described a process for development of corrective actions for resolving HEDs, the major weakness in the selection activity and the DCRDR as a whole is the high number of HEDs that rernain unresolved. Both the Summary Report and the audit discussions indicate u.any studies, evaluations, and reviews designed to resolve HEDs have yet to be accomplished. The results of these studies and reviews developed as a function of AP&L's design change process should provide HED solutions 18

i developed in accordance with good human factors engineering principles that are approved by the DCRDR team.

The results of these studies and evaluations will be combined into design packages that incorporate a cohesive set of corrective actions involving not only Category I HEDs but, potentially, Categories '2 and 3 as well if they are part of the overall corrective action. Of the 42 Category 1 NEDs that were identified in the Summary Report (Table 8.2, page 8-4),18 corrective actions have already been implemented in refueling outage IR6 (Nov. 84 - Jan. 85). The NRC audit team reviewed the corrective actions ,

implemented in the ANO-1 control room and found that these 18 HED solutions provided the appropriate correction of the deficiencies without creating any unacceptable human engineering discrepancies.

The selection of design improvements is incorporated into the Design i Process Procedure (AP&L procedure No. 202) that normally is required of any

- change to plant systems being considered by AP&L This process begins with an Engir eering Action Request (EAR) which 's written by the DCRDR team with 4 AP&L eng.neering support as necessary. 1he EAR is a request for preparation of a design change package or a revision / amendment to an existing design change package. A project engineer will review the EAR and any attachments

to scope the effort required to prepare the design change. If a Project Scoping Report (PSR) is necessary, then the Project Engineer will coordinate writing of the PSR with input from the Discipline Engineers to include Human Factors iiecialists.

The PSR serves to inform management of the design approach being considered, and the requirements of the approach, before detailed engineer-ing tegins. The PSR is to be used only as a project coordination tool and is c:,t used as design input. The PSR is reviewed by an HFS along with the Plant Operations Superintendent and the Plant Engineering Superintendent, and concurrence is required prior to developing a Design Change Package (DCP). The final authority for initiating a DCP is the Vice President of Nuclear Operations who will initiate a new EAR, which includes the PSR, requesting the preparation of a Design Change Package.

In conclusion, the licensee has developed a process for correcting and implementing improvements in the control room which is effective, judging r

19

--_-J__----.-.---_..-.--... -

from a review of the 18 HEDs already implemented as part of 1R6. This 4 process indicates AP&L's awareness of the need for implementation of corrective actions in an integrated fashion. However, the licensee has not presented adequate descriptions of design' improvements for each HED identi- ,

l fied in the Summary Report as required by Supplement 1 to NUREG-0737.

l Presently, ANO-1 has a-number of HEDs that are still unresolved awaiting i further review and selection of HED solutions as part of the Design Process j Procedure ( AP&L Procedure No. 202). The licensee must document all solu-l tions of Category 1 NEDs and Categories 2 and 3 HEDs if they become part of i

}

a design change package. The licensee must also provide justifications for  !

4 not correcting or partially correcting HEDs with safety significance. This ,

1ssue will remain an open item until the NRC reviews the proposed design l'

change solutions, implementation schedules, and justifications for safety-l related HEDs that will not be corrected or only partially corrected.

3. Verification That Selected Design Improvements Will Provide the Neces-i, sary Correction and Can Be Introduced in the Control Room Without j Creating Any Unacceptable Human Engineering Discrepancies While the licensee has not provided a formal process for the ,

. verification of the effectiveness of corrective actions, it appears that l

{ such a process does exist as part of the normal AP&L design process .

[

j procedure. i i <

l As described in the previous section each DCP will be evaluated with respect to the human factors aspect of any control room-related design ,

i changes ( AP&L Procedure No. 202 pages 22 and 23). The Project / Discipline Engineer in charge of the DCP will determine whether a human factors review j l 1s required. The human factors review, when required, will be performed as i j directed by the IAC Section Procedures. A human factors review checklist  :

I will be completed using an Information Request Form (IRF) detailing the j human factors review, and forwarded to the requesting engineer for inclusion  !

{ into the DCP. Only one human factors review checkItst is required per DCP.

The licensee indicated during the audit that verification of the l effectiveness of HED resolutions might be performed by using panel mock ups, j the simulator, and the involvement of Human Factors Specialists. Photo-l

. 20

. l a

graphs, blackboard walk-throughs, and panel blueline drawings were presented as techniques that are currently being employed.

In conclusion, the licensee has not finalized a formal' process for accomplishing this effort. However, a human factors review will be performed for control room-related design changes. Additionafly. AP&L is planning the use of various techniques such as mock-ups and blueline drawings for assessing the effectiveness of the HED resolutions. If the  :

processes as discussed at the audit are included in this effort, this process will meet the requirements of NUREG-0737, supplement 1. l

4. Coordination of DCRDR Improvements With Other NUREG-0737, Supplement 1 Improvement Programs ,

The Summary Report states that AP&L has a coordinated program for the implementation of NUREG-0737, supplement 1 initiatives which is intended to '

optimize the interface between the various initiatives. The organization t for the effectuation of the program, the relationships between the various initiatives (i.e. DCRDR, SPDS, Reg. Guide 1.g7, symptom-oriented E0Ps, and ERF), and the dates for completion of the various program milestones are ,

shown on a figure referred to but not included in the Summary Report. A copy of the referenced figure has been obtained and appears to demonstrate that a coordination program is established and functioning. .

I

.. i I

During the audit, the Audit Team attended a formal presentation on the coordination program and has concluded that this DCRDR element is being  ;

satisfactorily performed and meets the requirements of Supplement 1 to NUREG-0737.

DESCRIPTION OF PROPOSED DESIGN CHANGES AND JUSTIFICATION FOR HEDs WITH l SAFETY SIGNIFICANCE TO BE LEFT UNCORRECTED OR PARTIALLY CORRECTED  :

l Licensees are required by Supplement 1 to NUREG 0731 to submit an outline of proposed design changes including their proposed schedules for implementation and a summary justification for HEDs with safety significance to be left uncorrected or partially corrected.

21

i *

, . l

1. Proposed Schedules for Implementing HED Corrections AP&L's proposed approach for implementing HED corrections was provided

! in the cover letter which accompanied the Summary Report when it was trans- I r

! mitted to the NRC. As cited in that letter, the proposed four-phased

! program emphasizes an orderly and integrated process for the selection and implementation of HED solutions. While this process appears to 6e a satis-factory method for resolving AM0-l's HEDs, there is a concern that this i proposed phased approach does not have a sense of urgency for completing the corrective actions and is not associated with formal completion dates.

l At the audit. AP&L management, the NRC team leader and project manager proposed an agreement on a formal schedule for the completion of HED cor-

! rective actions. The formal agreement will be submitted by AP&L to the NRC  !

! within 30 days after the completion of the audit (September 19, 1985). ' l j Within six months after the 30-day agreement letter AP&L will submit to the NRC a supplement to the Summary Report. This supplement will include all  ;

proposed HED corrective actions, and justification for any HEDs left uncor-rected. In addition, a schedule for implementing these HED rescLlutions '

l within the next two refueling outages (1R7 and 1R8) will be provided. Any i corrective actions that can not be implemented by the second refueling ,

outage (1R8) will be documented along with a justification for the delay.  ;

I 1 "

2. Proposed Corrective Actions and Justifications for HEDs to be Left Uncorrected

{ .

I HEDs identified during the DCRDR are presented in Volume 2 of the  !

Summary Report. Findings are presented by HED categories:

l o Category 1 - 42 HEDs e Category 2 - 51 HEDs e Category 3 - 369 HEDs 4

dxamination of Volume 2 of the Summary Report indicated that several of the j Category 3 HEDs were not included. When notified of this omission, AP&L was able to supply the missing material at the audit.  ;

i Appendix A of this report contains a listing of HEDs for which proposed

~

22 i  :

e corrective actions, or justification for not correcting, were found to be inadequate. Four samples, with selected HEDs of the kinds of ambiguities which are causing concern regarding AP&L's plans / capabilities for adequate disposition of HEDs are presented below.

1. There are Category 1 NEDs that have been partially corr'ected using interim solutions without permanent resolutions having been de firied.

HED QS:A3.5-1.042: The discrepancy described is the need for a core exit thermocouple indicator in the control room. AP& L's response states that such an indication has been provided on the SPDS. While the SPDS provides an interim non-class 1E solution for

. this problem, a better resolution needs to be developed and imple mented.

2. There are Category 1 NEDs for which the licensee has not committed to implement proposed corrective actions until further evaluation.

Final disposition of these HEDs should be documented.

HED QS:A1.7 1.018: The HED described here is the need for remote control operation of the decay heat pump suction valves. AP& L's response indicates they Eg install a motor operator to the subject manual valves. While the proposed corrective action appears to be satisfactory, there is a concern that AP&L has not formally com-mitted to proceed with this corrective action for this safety- ,

significant (Category 1) HED.

'n 3. There are Category 1 NEDs which require further review and/or study before a decision can be reached as to whether or how the HED can be corrected or whether no correction can be justified.

HED QS:83.17 1.071: This discrepancy indicates that the service water instrumentation is not well laid out. AP&L indicates they intend to perform an evaluation in order to determine a corrective action. Untti a final solution is reviewed by the NRC staff. HEDs awaiting the results of studies or , reviews will remain open.

23 O

4. The . audit team is concerned with AP&L's position regarding certain HEDs.

HEDs CK:1-1.007 and CK:1-1.006: The discrepancies described here ,

are the location of 23 displays and 98 controls outside of the NUREG-0700-recommended evelopes. AP&L's response is that these deviations from NUREG-0700 do not create a significant problem for the control room operator. As a result, there are no plans for correcting these discrepancies at this time. This response does '

not identify the controls and displays affected or adequately

,. address the problems created by the location of the controls and displays covered by these two HEDs. It is recommended that AP&L identify the affected controls and displays and provide a justift- '

cation for not correcting them that addresses behavioral or opera-tional factors and issues related to plant safety.

CONCLUSIONS AND RECOMENDATIONS Arkansss Power and Light (AP&L) Company's Summary Report for the DCRDR conducted at Arkansas Nuclear One. Unit 1 (ANO-1) demonstrates a strong commitment towards meeting the requirements of NUREG-0737. Supplement 1.

The documentation submitted, in addition to extensive discussions of the review activities conducted to perform a DCRDR. indicates that AP&L basically met most of the requirements. However, additional information is re' quired from the licensee to provide assurances that all requirements as stated in NUREG-0737. Supplement I are satisfied.

The following is a summary of comments on AP&L's compliance with each of the DCRDR review steps and requirements, documented by the Summary Report and confirmed during discussion at the pre-implementation audit. The review portion was comprehensive, with the task analysis being conducted early enough in the process to become a key factor during the remaining phases of the DCROR.

e AP&L established a well-qualified, adequately staffed DCRDR team, which was composed of a good skill mixture to conduct the DCRDR.

Information relevant to levels of effort and staffing on DCRDR 24 -

_- . .- -. .- __ . -_ - - - . - ___.~ ---- .

i l

I .

1 tasks was provided at the pre-implementation audit. This require-ment of NUREG-0737 is satisfied.

? i .

! e Although not a requirement to Supplement 1 to NUREG-0737. s review-

! of operating experience was conducted consistent with NUREG-0700 i guidelines and objectives.

j e AP&L described a system function and task analysts based on the ,

j generic ATOGs which were made plant-specific. All unique tasks l were identified and broken down into task elements. Instrument and 1

. control requirements and relevant characteristics were identif,ted i for task elements. The methodology appears comprehensive and '

systematic. While there was a preliminary concern about the '

independence of the task analysis from the existing control room.

it is apparent from the audit that the existing instrumentation did

not bias the process. Instrument and control requirements w'ere 1 l

developed and subsequently checked using an iterative process  !

resulting in numerous HEDs. which is indicative of a properly f l'

executed process. The systen function and task analysis satisfac- i l torily meets the requirements of Supplement 1 to NUREG-0737.

1 ,

j e The licensee compiled a complete and comprehensive control room i inventory. A verification of equipment availability and  ;

suitability was then conducted by comparing information and control l .

I requirements determined from the task analysis with the equipment l l present in the control room as identified by the inventory. With j assurance that the information and control requirements were ,

! derived from a well-executed task analysis, it has been determined l that this comparison satisfactorily meets the requirements of Supplement I to NUREG-0737.

! , e The DCRDR documentation management system, which was automated, proved valuable and well-used in all phases of the DCROR.

i i e A human factors survey of the control room was conducted in what l appears to be a comprehensive and thorough manner. The methodology

! and objectives of the survey were es'sentf ally in accordance with the guidance provided in NUREG-0700 and met the requirement of ,

l l

1 l 25

, , . . , _ _ ,-_,,--.-.w- ..,,w,~ e.,v.-v, -- p , em wn, .-y w wy..-.,.,%.e, _ ce..,ew.,,.~,#w,my,n - . ., c er %,.w,,-..,w,,,.,~,_em,--,,

i .* ,

( l f Supplement 1 to NUREG-0737. AP&L's deviations from NUREG-0700 in f

conjunction with applicable justifications for such deviations were 4 submitted and discussed at the pre-implementation audit. As a f result, six differences between NUREG-0700 and AP&L's checklist were found unacceptable by the Audit Team. AP&L should' amend its i checklist to reflect the concerns and should reevaluate the control room in the areas in which the Audit Team disagrees with the AP&L l

checklist. Those checklist items found incomplete or discrepant should also be reevaluated.

e The validation approach implemented by the licensee was found to be .

i consistent with the guidelines of NUREG-0700.

{ e The process AP&L developed to assess the significance of HEDs appears to meet the requirements of Suppilement 1 to NUREG-0737.

l The HED assessment / categorization process resulted in the identift-l cation of 42 Category 1, 51 Category 2, and 369 Category 3.HEDs. l l All HEDs were reviewed for appropriateness of categorization by i SAIC' during the evaluation of the Summary Report. During the i

audit, all HEDs in Categories 1 and 2 and a representative sampling of HEDs in Category 3 were again reviewed. The c~onclusion reached

! was that there was no need for any category upgrades and that the HED assessment element of the DCRCR had been satisfactorily per-I formed.

j e While the process developed by AP&L to select design improvements 1s satisfactory, many HEDs are still unresolved awaiting further l ,

j review to develop the design solutions. The licensee must document l all solutions of Category 1 HEDs along with any Category 2 and

) Category 3 HEDs if they become part of a design change package or i have a cumulative or interactive effort raising them to a Category

! 1. All safety-significant HEDs left partially or completely uncor-l rected must be justified. Until these resolutions or improvements can be detailed, this requirement of NUREG-0737, Supplement 1, will-3l remain an open item.

e AP&L has not described a formal verification process to ensure that l

selected design improvements wi11 provide the necessary correction I

! 26 l

! i E_ _ _ . _ _ _ _ , _ _ ._ _

. ~ . . . --. _-. -. _ .- . .

without introducting new HEDs. However, it appears that such a l process does exist as part of the utility's normal design process change procedure. Additionally, the DCRDR review team may use mock-ups, the AN0-1 simulator, and blue line drawings and other techniques tgaccomplish the task of verification of HE'D resolu-tions. If APil follows a methodology that includes the above techniques, in conjunction with reviews by an HFS. it 'would appear that the licensee will meet this requirement to NUREG-0737, Supplement 1.

e The additional information provided by the licensee during the audit indicates that it is meeting the requirement to coordinate control room improvements with changes resulting from other improvement programs. ,

In addition to these general comments, the following is a list of the

activities, areas of improvement, and documentation that AP&L should satis- -

)

factorily perform in order to meet the NUREG-0737. Supplement I requirements for a DCRDR. It is recommended that this information be documented in a supplement to the Summaiy Report.

1. Control Room Survey e Modify the six AP&L checklist guidelines discussed at the audit ,

and restated in this report so they will be in accordance with the criteria of NUREG-0700. These revised guidelines should then be incorporated into AP&L's review process and applied to the ANO-l's control room review.

e During the audit, it was noted- that a number of survey checklist items were not completed. It is recommened that the utility resurvey these items and ir. corporate any findings into the review process, especially those related to ANO-l's color-coding used in the control room.

i 27 1

L _ _- . . _ . _ . _ . __. ._ .__ _. . . _ . . _ _._._ _ _ ,

2. Selection of Design Improvements l e Complete the HED studies and reviews that are outstanding.

Provide the NRC with a description of the proposed design i changes that will' result from these studies and how these j

results will be used to correct the control room dis,crepancies.

e Ensure that related HEDs in Categories 2 and 3 are considered l

for cum'ulative and interactive effects when resolving HEDs in Category 1 (e.g., annunciator system HEDs).

3. Verification of HED Resolutions e Formalize the process for this requirement detailing the use of personnel, equipment, procedures, or techniques that will ensure the satisfactory completion of this requirement, as AP&L's DCP does not go into great detail on how the HF review t will provide for the proper resolution of HEDs. I

4. Additional Activities Identified During the Audit e Six months after receipt of the 30-day letter, the supplement  ;

to be submitted will provide resolutions for those HEDs that are to be implemented by the IR7 and IR8 refueling outages, respectively. In those cases where provision of this infor-matiori may not be possible, , justification will be provided.

e Reexamine the color-coding scheme and provide NRC with a matrix of color versus meaning wherein color is used as a code.

e Reexamine the high location of the meters on panels C14,16 and i 18 (e.g., decay heat meters) with an eye toward their re-location so that they may be read conveniently from the control ,

room floor level. Use of a moveable ladder as a permanent solution in unsatisfactory.

l t .

28

e e Provide a procedure for testing status lights in cases involving single bulb lights.

I e Immediately ensure that there are no missing labels by using dymo-tape as an interim solution until the labeling study is implemented in IR7.

t e

e e g e

e O V e

% [4 s

9 29

_ _ , -_. ._. . . . . _ _ . . _ . . _ . . . _ , . _ . . _ . , _ . ~ . . . . _ . . . .

REFERENCES

1. " Arkansas Power and Light, Arkansas Nuclear One - Unit 1 Control Room Design Review Final Summary Report Volumes I & II," Arkansas. Power and Light Company, August 14, 1985.

t~

2. " Control Room Design Review Program Plan for Arkansas Nuclear One -

Units 1 & 2 Arkansas Power and Light Company," November 25, 1983.

3. "NRC Response to Arkansas Nuclear One Program Plant Submittal," USNRC,

. Washington, D.C., February 2, 1984.

4. NUREG-0660 Vol.1. "NRC Action Plan Developed as a Result of the TMI-2 Accident," USNRC, Washington, D.C., May 1980; Rev. 1 August 1980. ,

1 5. NUREG-0737, " Requirements for Emergency Response Capability," USNRC, Washington, D.C., November 1980.

6. NUREG-0737 Supplement 1 " Requirements for Emergen.cy Response ,

, Capability," USNRC, Washington, D.C., December 1982, transmitted to reactor licensees via Generic Letter 82-33 December 17, 1982.

7. NUREG-0700, " Guidelines for Control Room Design Reviews " USNRC, Washington, D.C., September 1981.

8. NUREG-0800, " Evaluation Criteria for Detailed Control Room Design Review," USNRC, October 1981.

e l

I s

30 .

l

. l d

APPENDIX A HEDs listed in Volume 2 of the Summary Report for which corrective actions I or justifications for not correcting were proposed. ,

1. The six HEDs listed below are Category 1 HEDs that are. partially corrected using interim solutions. Permanent resolutions for these HEDs should be finalized.

HED NUMBER SECTION-PAGE

~

.- CK:8-1.058 1-9 QS:A-1.17-1.001 1 - 11 QS:A-3.5-1.042 1 - 21 QS:1-1.097 1 - 31 QS:1-1.103 1 - 33 VR:1-1.006 1 - 34

2. The six HEDs listed below are Category 1 HEDs for which the licensee has not committed to implement proposed corrective ac'tions until

'further evaluation. Final disposition of these HEDs should be documented in the Supplement to the Summary Report.

HED NUMBER SECTION-PAGE QS:A1.7-1.018 1 - 13 QS:A1.8-1.019 1 - 14 QS:A1.9-1.031 1 - 19 QS:E2.2-1.084 1 - 30 VR:1-1.013 1 - 36 VR:1-1/031 1 - 40 c

31

3. No proposed solutions or justifications for not correcting are provided as the HEDs listed below are undergoing study or additional review.

HED NUMBER SECTION-PAGE .

i QS:B3.17-1.071 1 - 27

• CK:4-1.003 2-7&8 CK:5-1.044 2 - 12 CK:8-1.039 2 - 17 CK:8-1.046 ' 2 - 18 CK:8-1.055 2 - 20

- CK:8-1.061 2 - 21 CK:9-1.007 2 - 22 QS:B8.7-1.003 2 - 23 QS:B8.2-1.006 2 - 24 l QS:B8.3-1.007 2 - 25 QS:A1.2-1.015 2 - 28 QS:A1.19-1.038 2 - 35 QS:A3.12-1.045 2 - 36 QS:A3.23-1.054 .

2 - 37 QS:A3.30-1.061 2 - 38 QS:B3.18-1.072 2 - 39 QS:B8.2-1.092 2 - 43 ,

QS:1-1.093 2 - 45 QS:1-1.098 2.- 46

' VR:1-1.017 2 - 48 VS:1-1.017 2 - 50 .

VL:1-1.001 2 - 51 VL:1-1.004 2 - 52 o HEDs CK:4-1.008, CK:5-1 007, QS:B8.4-1.008 Presently, operator confusion can result from the f'ailure of single bulb status lights. Apparently ANO-1 operators can use independent indicators to determine the status of questionable indicators; however, AP&L needs to provide a means that will make the operators aware of any t

32

-,ww-., ry --weeWwm-v--r*e,,g e-m>=yme,gg-=e---e,+ue-e-

4 0 failed bulbs or indicator lights immediately and so that corrective actions. can be taken before the failure produces a safety-significant problem.

• e HEDs CK:7-1.04] CK:7-1.113 CK:7-1.048 CK:7-1.119 i

CK:7-1.049 CK:7-1.120 CK:7-1.050 CK:7-1.121 CK:7-1.051 The nine HEDs above related to the use of color on the SPOS and GERMS displays are not consistent with the rest of the control room.

Furthermore, color usage within these two systems is not standardized.

These inconsistent uses of coloring in the control room reduce oppor-tunities to employ the color as ~a visual cue and may lead to operator confusion.

AP&L was requested in the NRC exit briefing to develop a matrix for the

- use of color as a code in the ANO-1 control room. This matrix should provide the relationship between the various colors and their associated meanings to establish whether AP&L has established a con-sensual coding schema in the control room.

e HED's CK:5-1.009 CK:8-1.017 CK:5-1.103 CK:8-1.028 CK:5-1.029 CK:8-1.030 CK:5-1.034 CK:8-1.053 CK:5-1.035 -

CK:8-1.058 CK:5-1.039 CK:9-1.005 CK:5-1.042 CK:9-1.008 CK:5-1.043 CK:9-1.009 CK:5-1.046 CK:9-1.010 The 18 HEDs above describe numerous problems associated with the labeling in the ANO-1 control room. Included are problems such as labeling inconsistencies, incorrect abbreviations, the use of dymo-tape

! 33 l

l l

l g ..

and missing labels. The' NRC understands that a program is underway to  ;

relabel the control room in an integrated fashion. Many of the HEDs listed do not contain corrective actions since ,they will be considered during this labeling program. While AP&L's relabeling program sho.1d be able to resolve the majority of ANO-l's labeling deficiencies, the NRC is concerned with the fact that many instruments are uglabeled at this time. Immediate action should be taken to ensure that all control room instruments are labeled. This situation cannot wait for the final resolutions and implementation of the relabeling program.

e HEDs CK:1-1.004 CK:3-1.018 CK:3-1.003 CK:3-1.019 ,

CK:3-1.020 CK:3-1.004 CK:3-1.007 CK:3-1.021 CK:3-1.003 CK:3-1.022

. CK:3-1.009 CK:3-1.023 CK:3-1.010 CK:3-1.024 CK:3-1.011 CK:3-1.025 CK:3-1.012 CK:3-1.026 CK:3-1.013 QS:C2.4-1.010 CK:3-1.014 QS:C3.1-1.027 CK:3-1.015 QS:C3.18-1.022 CK:3-1.017 VR:1-1.004 The 26 HEDs listed above describe deficiencies related to the annuncia-tor system. The Summary Report describes- an annunciator upgrade I

program that is designed to address annunciator-related HEDs and a general upgrading of the annunciator system. Approximately half of the HEDs listed above will be evaluated by this program to determine the corrective action to be taken. Other HEDs should be recognized in light of the results of the annunciator study for cumulative or inter-I active effects. Of particular concern to the NRC is the present cap-ability to acknowledge the annunciators from any work station. As a result, annunciators can be acknowledged from a distance where the tiles are not readable. The NRC disagrees with AP&L's position that a this arrangement is appropriate for the AND-1 control room. The NRC 34 i

takes the position that the operator should be able to read all the annunciator tiles from the position at the work station where the annunciator acknowledge control for those tiles is located (NUREG-0700, Guideline 6.3.3.5). ,

i'

j. -

~..

l t

e S

i e

a 35 1

l - . - _ . _... , ._., _ _ _ _ _ ,

APPENDIX B ANO PRE-IMPLEMENTATION AUDIT MEETING SEPTEMBER 16, 1985 -

Richard J. Eckenrode NRC/DHFS/HFEB NRC Audit Team Leader.j-James M. Levine AP&L ANO General manager Stephen L. McKissick AP&L I&C Supervisor-CRDR Team Leader ,

Robert L. Kershner ARD Corp. Lead Human Factors Specialist CRDR Team Member Guy S. Vissing NRC/DL/ ORB 4 Project Manager, Unit 1 1 P. Harrell RI/AN0/NRC Resident Inspector Robert Lee NRC/DL/0RB3 Project Manager, Unit 2 Joe Moyer NRC/SAIC Human Factors Specialist James Hoyt ,

NRC/SAIC Human Factors Specialist, Raymond Roland NRC/SAIC Engineering

. Mark I. Good NRC/COMEX Operations W.D. Johnson NRC Senior Resident Insp ector James McWilliams AP&L NUREG-0737 Steering Committee Member B.A. Baker AP&L NUREG-0737 Steering Committee Member R.P. Wewers AN0/AP&L WCC Manager G.D. Provencher AN0/AP&L Gary G. Young UESC/AP&L CRDR Team Coordinator Daniel Williams AP&L Engineering Nuclear Service CRDR Team Member

$ A.J. Wrape AP&L E.E. Supervisor r CRDR Team Member B. A. Terwilliger AP&L Operations Assessment Supervisor CRDR Team Member

D. Kent Barnes ARD Corp. Systems Technology Group Bill L. Garrison AP&L Operations Technical Support CRDR Team Member Curtis W. Taylor AP&L Operations Technical Support CRDR Team Member 4

Patti Campbell AP&L l

36 ,

-- . . . ~ , - . _ - - , - . _ , _ . . . , _ _ _ _ _ , , _ , , _ . _ , _ _ _ , _ _ . _ _ _ _ . . _ __ _ _ _ _

%: *~ '

AND PRE-IMPLEMENTATION AUDIT MEETING SEPTEMBER 17, 1985 .

Richard J. Eckenrode NRC/DHFS/HFEB NRC' Audit Team Leader .

Stephen L. McKissick AP&L I&C Supervisor CRDR Team Leader '

Joe Moyer NRC/SAIC Human Factors Specialist James Hoyt NRC/SAIC Human Factors Specialist Raymond Roland NRC/SAIC Engineering Mark I. Good NRC/COMEX Operations Gary G. Young UESC/AP&L CRDR Team Coordinator

• . Robert Lee NRC/DL/ ORB 3 Project Manager, Unit 2 Robert L. Kershner ARD Corp. Lead Human Factors Specialist CRDR Team Member Daniel Williams AP&L Engineering Nuclear Service CRDR Team Member B.A. Terwilliger AP&L Operations Assessment Supervisor CRDR Team Member Bill L. Garrison AP&L Ope' rations Technical Support CRDR Team Member Curtis W. Taylor AP&L Operations Technical Support CRDR Team Member D. Kent Barnes ARD Corp. Systems Technology Group Don Taylor ARD Corp.

4 ,

1 37 i

l s l

ANO PRE-IMPLEMENTATION AUDIT MEETING SEPTEMBER 18 AND 19,1985*

i Richard J. Eckenrode NRC/DHFS/HFEB NRC Audit Team Leader .

Stephen L. McKissick AP&L I&C Supervisor-CRDR Team Leader ,..

Robert L. Kershner - ARD Corp. Lead Human Factors Specialist CRDR Team Member Joe Moyer NRC/SAIC Human Factors Specialist James Hoyt NRC/SAIC Human Factors Specialist Raymond Roland NRC/SAIC Engineering -

Mark I. Good NRC/COMEX Operations 2 Gary G. Young UESC/AP&L CRDR Team Coordinator

Daniel Williams AP&L Engineering Nuclear Service ,

CRDR Team Member B.A. Terwilliger AP&L Operations Assessment Supervisor CRDR Team Member Bill L. Garrison AP&L Operations Technical Support CRDR Team Member Curtis W. Taylor AP&L . ,0perations Technical Support CRDR Team Member l

! Cynthia Parr ARD Corp.

• Attendance was not taken on the 18th and 19th, but the above-listed j ,

persons are known to have comprised a portion of those in attendance.

~

O i

1 38 L _. _. ___-._._., ... _ .-_._._ _...,__ .___ .~.__._._...._..,._... _ _ .__ _ ._ _ ._ ... _ .

ANO PRE-IMPLEMENTATION AUDIT MEETING SEPTEMBER 20, 1985 Richard J. Eckenrode , NRC/DHFS/HFEB NRC Audit Team Leader -

James M. Levine i AP&L ANO General manager Stephen L. McKissick AP&L I&C Supervisor-CRDR j Team Leader Robert L. Kershner ARD Corp. Lead Human Factors Specialist CRDR Team Member ,

Guy S. Vissing NRC/DL/ ORB 4 Project Manager, Unit 1 P. Harrell RI/AN0/NRC Resident Inspector Robert Lee NRC/DL/ ORB 3 Project Manager, Unit 2 Joe Moyer NRC/SAIC Human Factors Specialist James Hoyt NRC/SAIC Human Factors Specialist Raymond Roland NRC/SAIC Engineering Mark I. Good NRC/COMEX Operations-Gary G. Young ' UESC/AP&L CRDR Team Coordinator

. Daniel Williams AP&L Engineering Nuclear Service CRDR Team Member A.J. Wrape AP&L E.E. Supervisor CRDR Team Member B.A. Terwilliger AP&L Operations Assessment Supervisor CRDR Team Member Bill L. Garrison AP&L Operations Technical Support .

CRDR Team Member Curtis W. Taylo'r' AP&L Operations Technical Support CRDR Team Member Don Lomax AP&L Plant Licensing Supervisor R.P. Wewers AP&L WCC Manager D. Kent Barnes ARD Corp. Systems Technology Group Jack Orlicek AP&L Field Construction Management Basil Baker AP&L Operations Manager M.L. Pendergrass AP&L Engineering and Technical Support Manager  !

L.W. Humphrey AP&L Administrative Manager L.W. Schempp AP&L Manager, Nuclear QC 39

l g; ...

APPERDIX C DIFFERENCES BETWEEN NUREG-0700 AND AP&L CHECKLIST PRESENTED AT ANO-1 PRE-IMPLEMENTATION AUDIT ON SEPTEMBER 18, 1985 i

• Indicates those items examined at the audit. j-

- I i

f.'i' Resolution of Differences Between *

'l 3 , 3rUREG-0700 section 6 and AFL Checklist,

. . . . . c. .

• Checklist 2 tem , .. - -

'2apact~*

Different From*n '

• en AFL' 7

{

• truREG-0700 '* *: *- '

' Resolution . i survey ~ - -

' , ,. ., .. .. .s ... . , , d'"

t . ,% ,* '. . "' 3 y.4= T .p. . 91 r

-;i:.e.,

wor dinghalue.;*,};

.1.1.1.a . c Change to WUREG-0700 yEy.**.h*i. t Change to IrDREG-0700 wordinghalue. -b 1.1.2.a)!.' VChange 'to trUREG-0700 'wordinghalueN

.', .b h , . , Change to NUREG-0700 wordinghalueN,'

1.1.3.e(1) ,, . . Change to,3rUREG-0700 wordinghal'ue.*.*p,

• 1.1.3.f(1) *" . -

",Change thangetotoBrUREG-0700

,3rUREG-0700 wordinghalue.N, (2) wordinghalue.

(3)

• l ' ' Change to 'IrUREG-0700 wordinghalue.' .

Exhibit 1-3 ', Change to 3rUREG-0700.wordinghalue.'l 1-4 . Chanye to 3rUREG-0700 wordinghalue.7 1.2.2.b(1) Change to trUREG-0700 wordinghalue-Add

, a page for Exhibit 6.1-6 and a space for drawing panel. * '.*'. * * ,, ,, , ,,

(2) Change to trUREG-0700 wordinghalue.

• Exhibit ~

6.1-6 Change 8700 to" extended reach value of

, 29" for 5th percentile" female. ' '

• 1.2. 2.d ( 2)[ Change value to 29" which is ace'e'pt$d extended functional reach for 5th , i percentile female. , .

f

• 1.2. 2. e( 2 ) 3rURIG-0700 suggests that the -

measurement for eye reference be taken * -

from the leading edge of the benchboard.

! ARD takes the stand that the operator

bas some maneuverability when reading s and his eye is closer to 4

' inches display %kfromthebenchboard.'*[*T' i Sir.ilarly, ARD suggests that the , ,

operators reference point for annunciators is 16 inches in lieu of .-

l the 12 inch nominal distance,provided as guidance in 3rUREG-0700. , *

• 1.2.3.c.

. The reach criteria has already been .

established in 1.2.3.b. The 0700 l criteria does not provide guidance for .

l an acceptable slope angle. The optimum angle of the benchboard surface would **

depend upon its use. If the surface 40 -

theck11et Item #

e..

Zapact.' '. l*.' Y :~'I . .

oiff. rent rrom . . ';!M - .- s- -

r-

, n m.-l,

' . ' - i : ",

WURIG-0700

• Resolution survey ** ",- .; ., ', l

,R . . .

. . . '. n . . r . * *

. contained a keyboard the angle should, -

be within 0-156 (van Cott) whereas'for - l viewing the optimum angle is 450 (Van

• l Cott) . McCormick provides desirable i ranges for the angle of the benchboard as 150-300 for wr'iting and typing surface and' 300-500 for benchboard contairiing primary controls and some *

, related displays. Since the, criteria i

3:.

.. refers to a benchboard and infers a '

d;.. g* -%. . * , * - .sicpe of some kind, a reasonable range

,- , is 150-450 (as shown in Exhibits .

5.1-6). The optimum would depend upon -

+.

the activity performed at the , . .

• p benehboard.,[,'g gh,,ff,,Q,w c..,p., ..[,i.*5

1. . . . . ..2.3.b lc. ..* see 1.2.2.d(2). . Change,value to 29". . ..;

i.

o

.[123d(2) es ...

. 1.2.3.e(2),

,

• See 1.2.2.d(2).'" Diange value to 29".

,, Delete *The upper limit is 56 inches".

2nclude Exhibit 6.1-10 and the words

~ * ' . . . .. ; . . . . * - (see Exhibit -

). * *

• 6

'1.2.3.f(2)

T see 1.2.2.d(2). Change value to 29",

1.2.5.a(1)

Change to NUREG-0700 wording /value.

1.2.5.b(1) Change to NURIG-070'0 Twerding/value.

(2) Change to NURIG-0700 wording /value.

! 1.2.8.d Change to NURIG-0700 wording /value.

4 1.5.3.a Oiange to NUREG-0700 wording /value.

j

• 1. 5. 3. b Creatly is too subjective of a term.

• Cur experience is that 15fc is a good

'value for this quantification.

• 1. 5. 5. a ( 2 ) . This value is again selected based upon our experience with noise measurements.

This is to quantify a subjective ~

requirement. .

. 2.1.1. b,

• Effective" cannot be measured during periodic testing.

2.1.1.c(2) "and are known to operators" is a redundant sta'tement. It is assumed that if procedures are in place, they are covered in training.

2.1.2.b(6) "by passing traffic" or by anyone, the concern is with knocking the phone out of the cradle.

2.1.2.d Change to NURIG-0700 wording /value.

3.1.2.a(1) Otange to NURIG-0700 wording /value.

3.1.2.b(1) Missing some words. Change to NURIG-0700 wording. , , ..

• 3.1. 2. c (1) Instead of using the word " avoided" you .

can use the multiple annunciator alaras if you have ser.e backup information.

See 3.1.2.c(2).

41 l

Checklist 2 tem . . '- - -

.. . C.' 1spect ,

5 ,, ,

Different From .', I,,* , , , ' en AP{. ..',,, ,

.i i.~,,

.<j-r NUREC-0700 Resolution - Survey - ,

• 3.1.4.b(1) Change to NUREG-0700 wording /value. ..

• 3.2.1.c 90 da(A) is the accepted threshold of ,

. pain for auditory signals. This is  ;

provided to quantify checklist item.

• 3.2.1.d Incremented steps in loudness are slightly noticeable at 2 da steps and fully discernable at 5 da. Nowever, .

since different frequencies are being produced'by the different alarms, the NRC did not specify a value. In order to evaluate the item ARD selected +2.5 a' which will result in approximately . *

• ~

. .....,' h,). equal sounding detection levels. -

.,,3'.' 2.1.f ;; , 1.ine skipped. Change.toi700.*.f;,,'2

• f [(

• MPN'S .',*.),,., wording /value.' ' eC.F**P.. f: . ' "

'3.3.1.a *

• Delete asterisk. Same as 0700.~F "

3.3.4.b ,..') Delete asterisk.' same as 0700 '

,4.1.1.a(2) Delete asterisk. Same as 0700 except easily is deleted. Easily is a .

subjective, not quantifiable term. '

• 4.1.1;d Easily is a subjective, not .

quantifiable ters.

4.2.1.e .

MIL STD 1472, the raference for the

.f guideline, makes no distinction between increase and raise and similarly between decrease and lower.

• 4. 2. 2. c ( 4 )

Type on thickness. Original source .

(McCorrdek) states that 3/8" in

, thickness can be identified very accurately by touch. These values were coni * 'ted to decimals in MIL STD 1472 and .Lc was more convenient and conservative for them to use .4. The

' experimental evidence actually suggests 3/2" as a liiit. -

4.2.2.e(1) Typo. Change to Exhibit 4-3.

(2) Typo. Change to Exhibit 4-4.

j (3) Typo. Change to Exhibit 4-5.

l' 4.3.1.a change tt 0700 wording /value.

• 4.3.2.a(1) Van Cott suggested a lower limit on pushbuttet. controls as 0.5. Nowever, MIL 8 2 1472 suggested 3/8" or 0.375".

0.385 has no significance and it is believed that it is a misprint in NURIG-0700 4.3.3.c(1) Change to 0700 wording /value.

(3) -

42

I g, ..-

Checklist Item , *

,. . - Impact ~- '

~ * * '

• Different From .7 . .

. , . en AFL .. ,U. y l Resolution Survey " ** - * **

, . / ' , -

WUREG-0700 -

• 4.4.1.b There are other types of coding other .

,than shape which could serve the same -

purpose.

• 4.4.3.g(1) 800 for the minimum was a typo in g

'9700. MIL STD 1472C states 300 for *,

this value. . l

• 4.4.4(c)

Wumb and finger encircled is emmitted because it is pra,ctically never used in nuclear power applications and the term 1 is very confusing without a diagram dem-onstr'ating what it is. The fingertip * * *

'- actually refers to the gererally accept'ed continuoas rotary control used in the muciear industry which is grasped between the thumb and the forefingers.

The thumb and finger encircled is .';

'largerbecauseitisgraspedusin[the

, thumb with the forefinger, wrapping..

, . around the cireur.ference of the knob *

• similar to grasping a door knob. W. .

• *l4. 5.1.d ( 2 ) change the values to reflect the gdated '

~

ones from MIL sto 1472 c. Affects min-imum diameter and maximum trough distance only. . . > .

i 4.5.3.c(2) change to 0700 wording /value. . .

l 4.5.4.a statement for 4.5.4.a(1) was changed from 0700 to provide guidance for ver-tical orientation. The statement in 4.5.4.a(2) provides additional guidance

. based on MIL MOBK 759 A when horisontal orientation is used. .

1 4.5.4.b(1) An integral light is best but should not I be the only acceptable light feedback for a rocker type switch. A separate l '

f light located adjacent to the switch .

would also suf fice.

• • 4.5. 4.e (1) . These values are the updated values free MIL 37D 1472 c.

(4) change to 0700 wording /value.

5.1.1 The nurlering scheee has been changed from 0700 to provide more logical grouping of ite..s.

5.1.2.f change to 0700 wording /value.

5.1.3.a Delete preferred visual angle, reference.

Not a checklist consideration.

• l *5.1.6.c.2(b) Added to provide industry standard not l acknowledged by 0700.

e> Delete "Arber (yellow): Auto trip". .

This is not an industry standard.

43

,_ .-w_, ,.__,__vw,__w.__%...,#,,_.._,_,,,_m,.m,-.. _ - - , , _ - - . _ , - . _ _ - - . . - -

checklist Item ,

2apact .- " .-

Different From ./ *

. . * - ' I-

' , , . en survey m ,; ,

3rDRIC-0700 -

Resolution .

1 5.1.6.e(2) Delete asterisk. Same as 0700 ..,..  ; .

5.2.2.a(2) Delete asterisk. same as 0700.' -

5.3.1.a(1) change to 0700 wording /value. -

3.3.2.b -

The intent of the criteria is to ensure j

, the light intensity is sufficient to accurately determine if the indicator is lit. A spot photometer will give a very accurate reading of the intensity, . I however a simple photometer provides an accurate enough measure to* meet the intent of the guideline.

1 5.3.3.a(1) See 5.3.2.b.

• 5.4.1 Typo. * *

. . u .

15.4.1.1 . Format (numbering scheme) shange. ,

,5.5.1.a(5) , A matte finish is one good way to

.'- minimise glare and it is q good design i' ?2 .

criteria to have in MIL s2D 1472, etc.

Ef*)O ... *;;. 7. , '.

' ' ' .Slowsver,*the inteat should be to eval ,.

este whether the surface is free from -

~

glare regardless of the type of finish.

~; ,

5.5.3.a ', typo. , , ,;; , *

• 6.1.1 This guideline refers to the physical presence of a label for every control, display, or other ,quipment.

6.2.1.a change to 0700 w,rding/value.

.c Otange to 0700 wording /value.

.e Omnge to 0700 wording /value. -

6.2.3.a(1) *and read from left to right" is redund- ,

ant to oriented horisontally.

Exhibit 6.3 . Grass green is not very descriptive.

Dark green is a more far.iliar statement to describe the intended color. -

• 6.5.1.g Tag cuts cannot physically prevent act-mation of a control. ghe best it can do is indicate which controls should not be actuated.

6.6 The " heed for Iccation Aids" section is tutorial and does not contain Tuideline checklist raterial.

• 6. 6. 3.b The fact that differential l'ine widths

,may be used to code flow is tuitorial.

There is nothing to say other methods cannot be used or that this is the pre-ferred method.

7.1.2.a(4) operators speak in terms of acronyms not (5) syntax as do coeputer operators.

44 i

- - _ - , -_.-_-_-.___---._,._.-._.___a____. _ _ _ . _ . _ _ _ . . . _ . _ .

0

• a mecklist 3 tem .

Ispect . . .

Different From -

., k en APL *' 9 * , . . , .

' Resolution 3ruREG-0700

' ' ~ f* -

Survey *

' .7

• 7.1.4. g Chk.ir, et al. in the VDT Manual recom- .

sends a 50-150 keyboard slope based -

on experimental evidence.

• Galitz in aluman Factors in Office Automation ree- l' ommends 100-150 and MIL STD 1472 C, 150-250 The 100-250 range is a good compromise range betwen the con-fileting doceents. *

• 7.1.8.b(1)

Change to 0700 wordinghalue.

Exhibit 7.2 Change to 0700 wordinghalue.

7.2.1.a Delete asterisk. Same as 0700.

7.2.1.ct3) Delete asterisk., same as 0700.

7.2.2.g(3) Change to 0700 wording /value.

7.2.4.n Change to 0700 wordinghalue. ,

7.3.1.f.(2) Typo. -

8.1.1.a change to 0700 wording /value.

3.1.1.b .-

Thelastpartofthe6700(geleMineis '

t .. -

'.. tutorial. . - -

• 8.i.2 %. **'

• Effective Panel Idyout" section is tut-

[; , * '

orial and'dE s not i t;,' . .*, y; ,4.,5,-4

[. ,;*.checklist material.,contain

• .. . y .. . . guideline"

,, . ,1, .. -

• gt. 2.1. a ( 3 ) . .synemetrical is too limited for t.his

.' '

• guideline.

• They are appropriate,* ** * *** *

's 1ogical patterns that are not necessar-11y symmetrical which could relate a '

-aet of controls to displays.

'8.3.2.b 'Inss than two^ inches wide is a reason-Jable definition of small displays.

• Added for quantification.

8.3.2.c(1) Change to 0700 wordinghalue.

(2) Change to 0700 wording /value.

8.3.2.6 Add 'I4rge matrices are subdivided by appropriate demarcation". .

9.1.2.e(1) Typo. .

j

• 9. 2. 2 Control / Display packages is dialeted l since modular construction not used j in typical Cp .

] +9.3.1.b(4) Added to give sufficient guidance for j display types.

] 9.3.1.c(1) " Apparent" added.because there must be a time lag, but it should not be sig-l nificant for operator feedback purposes.

, 9.3.1.c(2) Change to 0700 wordinghalue.

9.3.2.a change to 0700 wording /value.

4346a I

i 45 A A

i APPENDIX D j A BRIEF DESCRIPTION OF REMOTE SHUTDOWN CAPABILITY AND COMMENTS 4 .

i The AP&L remote shutdown capability was not included in the DCRDR survey. However, two members of the NRC audit team had a walk-through with a Senior Licensed Operator using the procedure to demonstrate hoir the plant would be shutdown in the event the control room became inaccessible. Unit 1 4

has two procedures for shutting down the plant from outside the control room. The remote shutdown procedure (1203.29 Revision 1) was stated to be l

used when required control room evacuation is caused by anything but fire or l some other emergency that might affect the cabling, circuits, and panels in I the control room and cable spreading rooms. The alternate shutdown proced-ure (1203.02 Revision 0) would be used in case of a fire causing control '

4 room evacuation. The alternate shutdown procedure is broken down into three

' sections. The first section has two parts. Section IA covers any fire in i

the control room or cable spread room which requires an immediate control I room evacuation. In this case it is assumed that sufficient time is

}~ available to take a minimum number of actions prior to the evacuation.

l Section IB covers fires in the these areas that do not require an immediate control room evacuation, but that allow a certain number of specified actions to be taken prior to evacuation which will make the transition in

plant control more orderly.Section II provides guidance for maintaining

] the plant at hot shutdown, reestablishing plant control from the control room, and reestablishing individual component control from the control room.

Section III covers the steps required to perform an alternate shutdown plant cooldown using safe rSutdown components. Sections IA or IB require four

-operators.

The licensed operator had miror difficulties in performing the alter-nate shutdown procedures due to procedural, labeling, and lighting deficien-cies. The ability to shutdown the plant in an emergency with the added stress and confusic.) could be greatly enhanced by several minor procedural /

]

administrative changes.

I 46 l l

l

. i

\

PROCEDURE 1203.02

1. Page 8 of 47 Note specifies breaker charging tools but does not specifically mention diesel generator governor props or 480V breaker closing tools as being required. This would entail a return to the locker to get necessary tools in the middle of the procedure, should the operator not remember what tools to take.

[ 2. Step 5, procedure names do not match breaker label names.

,i

3. Step 7, label missing on panel 00-1.

,! 4 Step 7. modification to panel since procedure was written; breaker handle missing on exterior panel; panel had to be opened to operate l breaker. -

5. Normal lighting in 'the area of DO-1 not bright enough for operator to l

read breaker labels.

6., At D-11 double number labels on breakers because panel was modified, double pole breakers replaced with single pole breakers with double

' pole breaker numbering not removed. In addition, breaker labeling legend on panel door listed wrong components for labeled breakers.

l 7. Panel D1116A is not labeled.

8. At panel D-21, labeling deficiencies.

I

9. Certain breakers require pushing the trip button to open; operators l'

generally do not stick their fingers in the hole to trip these breakers. A portion of a broom handle or other nonconductive device i

might be appropriate to clip to the side of a switchboard or have available .to assist the operators.

i The format of Section 1203.02 Section IB is different from the format

of Section I A. Because it is a procedure that may never or seldom be used, operators may be less familiar with it than other procedures and it might be appropriate to eliminate potential sources of confusion.

! 47 r

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