ML20237H069

Rev 1 to Technical Evaluation Rept of Dcrdr for Sequoyah Nuclear Power Plant Units 1 & 2

ML20237H069
Person / Time
Site:	Sequoyah
Issue date:	08/10/1987
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML20237H072	List: ML20237H069
References
CON-NRC-03-82-096, CON-NRC-3-82-96 SAIC-87-3048, SAIC-87-3048-R01, SAIC-87-3048-R1, TAC-51203, TAC-51204, NUDOCS 8708170038
Download: ML20237H069 (142)

Text

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t-ENCLOSURE 2

' SAIC/87-3048

Revision 1 TECHNICAL EVALUATION REPORT

- 0F THE DETAILED ~ CONTROL ROOM DESIGN REVIEW FOR-TENNESSEE VALLEY AUTHORITY'S SEQUOYAH NUCLEAR POWER PLANT UNITS 1 AND 2 TAC N0s. 51203, 51204

,1 August 10, 1987 5AIC Scence AmIncatons knematonalCorporatoon i.

i Prepared for:

U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Contract NRC-03-82-096 37M74# N Post Office Box 13tD 1710 Goodridge Drive, McLean, Virginia 221M, t7@ W1-4300

f-TABLE OF CONTENTS Section EiLqe I

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . 1

2. 0 EVALUATION . . . . . . . . . . . . . . . . . . . . . . . 3 2.1 Establishment of a Qualified Multidisciplinary Review Team . . . . . . . . . . . . . . . . . . . . 3 2.2 System Function and Task Analysis . . . . . . . . . 4 2.3 Comparison of Display and Control Requirements With a Control Rocm Inventory . . . . . . . . . . . . 6 2.4 Control Room Survey . . . . . . . . . . . . . . . . 7 2.5 Assessment of Human Engineering Discrepancies (HEDs) to Determine Which Are Significant and Should Be Corrected . . . . . . . . . . . . . . . . 8 2.6 Selection of Design Improvemei.;s ......... 13 2.7 Verification That Selected Design Improvements Will Provide the Necessary Correction ......... 18-2.8 Verification That the Selected Design Improvements  !

Will not Introduce New HEDs . . . . . . . . . . . . 19 1

2.9 Coordination of Control Room Improvements With i Changes From Other Improvement Programs Such'as the  !

Safety Parameter Display System, Operator Training,

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Reg. Guide 1.97 Instrumentation, and Upgraded i Emergency Operating Procedures .......... 19

3.0 CONCLUSION

S ...................... '20 l

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4.0 CONFIRMATORY DOCUMENTATION NEEDED ........... 23

5.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . 25 ATTACHMENT 1 - List of Meeting Attendees ATTACHMENT 2 - Meeting Agenda and HED Questions ATTACHMENT 3 - TVA Meeting Presentation Slides ATTACHMENT 4 - Procedures Evaluated During Task Analysis ATTACHMENT 5 - Sample Task Analysis Documentation i

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l TABLE OF CONTENTS'(Continued)

Section ATTACHMENT 6 - Audit Team Sample Survey Human Engineering Discrepancies ATTACHMENT 7 - Assessment Procedures ATTACHMENT 8 - Category 1 and 2 HED Corrective Actions ATTACHMENT 9 - November 26,.1986 SQN-EOP/CRDR Meeting ATTACHMENT 10 - February 9,,1987 SQN DCRDR Training Related HECs -

ATTACHMENT'11 - Watts Bar/Sequoyah HEC Comparison Form ATTACHMENT 12 - March 13, 1987 Restart Requirement Criteria ATTACHMENT 13 Revised TVA Commitment Regarding HED 0220 ATTACHMENT 14 - TVA Employee Concerns Evaluation o

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l' TECHNICAL EVALUATION REPORT 0F THE 1 DETAILED CONTROL ROOM DESIGN REVIEW FOR l'

TENNESSEE VALLEY AUTHORITY'S SEQUOYAH NUCLEAR POWER PLANT UNITS 1 AND 2

1.0 INTRODUCTION

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The Tennessee Valley Authority (TVA) submitted a' generic Detailed )

Control Room Design Review (DCRDR) Program Plan to the Nuclear Regulatory Commission (NRC) on June 9, 1983 (Reference 1) in order to satisfy the q Program Plan requirements of Supplement I to NUREG-0737 (Reference 2) for l the Sequoyah, Watts Bar, Bellefonte and Browns Ferry Nuclear Plants. The Program Plan was resubmitted September 13, 1983-(Reference.3) to correct ]

l duplicating errors in the original plan. The NRC staff reviewed the {

submittal with reference to the nine DCRDR requirements of Supplement I to l I

NUREG-0737, and the guidance provided in NUREG-0700 (Reference'4) and draft NUREG-0801 (Reference 5).

Supplement I to NUREG-0737 requires that a Program Plan be submitted within two months of the start of - the DCRDR. Consistent with the )

requirements of Supplement 1 to NUREG-0737, the Program Plan should describe how the folicwing elements of the DCRDR will be accomplished:

1. Establishment of a qualified multidisciplinary review team

2. 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 identjfy deviations from accepted human factors principles. l 1

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

6. Selection of design improvements.

7. Verification that selected design improvements will provide the l 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 SPDS, operator training, Reg. Guide 1.97 instrumentation, and upgraded emergency operating procedures.

The staff comments on TVA DCRDR Program Plan review were forwarded to TVA by letter dated November 17, 1983 (Reference 6). Based on the Program Pl an review, the staff concluded that TVA addressed most of the nine requirements of a DCRDR specified in Supplement I to NUREG-0737. However, the staff determined that certain elements, notably the task analysis, should be strengthened to provide reasonable assurance that the control room reviews based on the plan will produce results that satisfy NRC requirements. In order to address the staff's Program Plan concerns, the NRC recommended that a meeting be held at NRC's Bethesda offices.

The meeting between NRC and TVA was held on June 14, 1984, in order to provide f.urther detailed information and address the staff's Program Plan review cancerns. As a result of this meeting, NRC indicated to TVA that an opportunity to more completely assess TVA's methodology for performing the system function and task analysis activity may involve an in-progress audit at Sequoyah. However, no in-progress audit was conducted at Sequoyah during the DCRDR.

Supplemenf 1 to NUREG-0737 requires that a Summary Report be submitted at the end of the DCRDR. As a minimum, it shall:

1. Outline proposed control room changes.

2. Outline proposed schedules for implementation.

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3. Provide summary justification for HEDs with safety significance to -

be left uncorrected or partially corrected.

T6nnessee Valley Authority (TVA) submitted a Summary Report for the Sequoyah Nuclear Power Plant Units 1 and 2 to the Nuclear Regulatory Commission (NRC) on October 31, 1986 (Reference 7). The Summary Report was l reviewed by Science Applications International Corporation (SAIC) personnel and a pre-implementation audit was conducted on June 22, 1987 through June 25, 1987. The audit team consisted of an NRC staff member, an SAIC repre-sentative, and a representative from Comex Corporation. Together, the team represented the disciplines of nuclear systems engineering, reactor opera-tions, and human factors engineering.

This Technical Evaluation Report reflects the consolidated observations, findings, and conclusions of the audit team members. A list of audit meeting attendees is provided in Attachment 1 and the audit agenda is provided in Attachment 2.

2.0 EVALUATION The purpose of the evaluation was to determine whether the nine DCRDR requirements in Supplement 1 to NUREG-0737 had been satisfied. The evaluation was performed by comparing the information provided by TVA with ,

the criteria in NUREG-0800, Section 18.1, Rev. O, Appendix A of the Standard Review Plan (Reference 8). The reviewers' evaluation of the DCRDR for the Sequoyah Nuclear Power Plant (SNPP), and a summary of the criteria from the Standard Review Plan are provided below.

l 2.1 Establishment of a Qualified Multidisciplinary 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 I be given sufficient authority to carry out its mission. A core group of specialists' in the fields of human factors engineering and nuclear engineering are expected to participate with assistance as required from ,

personnel in other disciplines. Staffing for each technical task should j 3

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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 briefing on DCRDR purpose and objectives which contributes'to the success of the DCRDR. NUREG-0800, Section 18-1, Appendix A describes criteria for the multidisciplinary review I team in more detail.

The overall administrative leadership of the DCRDR team was provided by a TVA employee who is presently also the assistant operations manager at the Sequoyah station. The TVA DCRDR administrator will continue to manage the project through the modification implementation phase. The Sequoyah DCRDR study team consisted of a core group of specialists in the fields of nuclear engineering, instrumentation and control engineering, reactor operations, and human factors engineering (see Attachment 3). The human factors engineering expertise was provided by an individual human factors specialist during 1984. Following a six-month break in Sequoyah DCRDR activities, Essex Corporation was contracted to provide human factors support. .Each TVA l DCRDR team member was given a two-day course in human factors engineering and control room design including the purpose and objectives of DCRDR.

The audit team evaluated the staffing for each technical task and determined that the appropriate expertise was present. In addition, the audit team determined that the DCRDR study team maintained the correct team disciplines, despite the six month break in Sequoyah DCRDR activity due to Watts Bar Nuclear Power Station DCRDR activities. It is the audit team's judgment that TVA has met the Supplement I to NUREG-0737 requirement for a qualified multidisciplinary review team.

2.2 System Function and Task Analysis The purpose of the system function and task analysis is to identify the control room operators' tasks during emergency operations and to determine the information and control capabilities the operators need in the control 4

room to perform those tasks. An acceptable process for conducting the function and task analysis is as follows:

1. Analyze the functions performed by systems in responding to transients and accidents in order to identify and describe those tasks operators are expected to perform.

2. For each task identified in Item 1 above, determine the information (e.g., parameter, value, status) which signals the need to perform the task, the control capabilities needed to perform the task, and the feedback information needed to monitor task performance.

3. Analyze the information and control capability needs identified in Item 2 above to determine appropriate characteristics for displays and controls to satisfy those needs.

The Sequoyah DCRDR task analysis methodology was presented in Section 2.0 of the Summary Report. Two separate task analysis efforts were con-ducted. TVA conducted two task analysis efforts in order to address previous staff concerns identified during the Program Plan review (Reference

6) and the June 14, 1984 TVA/NRC DCRDR meeting in Bethesda, Maryland. The first effort, called the Integrated Task Analysis, was a combined effort between the DCRDR team and an SNPP-staffed procedures team that developed the upgraded (symptomatic) Emergency Operating Procedures. The second effort, termed the Supplemental Task Analysis, was performed by the SNPP DCRDR team and a human factors consultant (Essex Corporation) to assure that information and control requirements were developed independently of existing control room equipment.

Both function and task analysis efforts were based on all of the site-specific emergency response guidelines, developed from the generic Westinghouse Owners' Group (WOG) Emergency Response Guidelines (ERGS), High Pressure version, Rev. 1, September 1983. Since Sequoyah has an ice condenser type containment, supplemental ERGS addressing this and other differences were also used. Therefore, differences between the generic and 5

plant-specific ERGS were considered. A list of emergency operating procedures that were analyzed during task analysis is provided in Attachment 4.

The audit team selected action steps from both the generic and supplemental ERGS and traced the methodology under which each of the task analysis methods were performed to determine the adequacy of the methods used and availability of documentation (see Attachment 5). It was noted that the sample set of tasks reviewed by the audit team, were thoroughly analyzed, including the alternate Response Not Obtained Column Tasks, Cautions, Warnings and Notes. In addition, documentation was adequate and j l was readily available and auditable. l i

Based upon the audit team's review of the methodology and the sampling l l

of DCRDR team activities, it is the team's judgment that the TVA (as l completed a system function and task analysis for the Sequoyah Nuclear Power i plant which was conducted independent of the existing control room and' meets' the requirement of Supplement I to NUREG 0737.

I 2.3 Comparison of Display and Control Requirements with a Control Room Inventory The purptse of comparing display and control requirements to a control room inventory is to determine the availability and suitability of displays l and controls required to perform the ERGS. The success of this element depends on the quafity of the function and task analysis and the control room inventory. The control room inventory should be a complete j l

representation of displays and controls currently in the control room. The i j inventory should include appropriate characteristics of current displays and l l

controls to allow meaningful comparison to the results of the funi: tion and  !

task analysis. Unavailable or unsuitable displays and controls should be I documented as human engineering discrepancies (HEDs). f For each of the two task analysis efforts discussed in Section 2.2 above, the " inventory" of instruments and controls used by Sequoyah was the l Control Room (and other panels as appropriate) itself. For the Supplemental l

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l Task Analysis, some additional plant documentation, such as plant diagrams, l were used in making availability and suitability determinations.

l The verification of instrument and control availability and suitability was accomplished by comparing the operator's requirements during emergency operations derived from the two task analysis activities to the equipment in I the Sequoyah control room. During the verification of instrumentation availability, the DCRDR team identified unavailable displays such as main j steam radiation monitors. During the verification of instrumentation suitability, the DCRDR team identified instrumentation and control suitability problems such as low pressure safety injection valves not functionally grouped together with the rest of the system. A " walk- and talk-through" by DCRDR team members and qualified operators was performed for each of the steps analyzed on the task analysis worksheets. " Human Factors Guidelines" checksheets (Attachment 5) were used to evaluate the adequacy of the instrument / control demonstrated by the operator, and the i information/ control equipment for #ulfilling the task analysis requirement.

Real-time simulations were also performed using highly time-dependent emergency procedures to evaluate perceptual-cognitive loading, communications, and spatial relationships. Potential HEDs were documented as human engineering concerns (HECs) during the review phase, and then converted to HEDs during the assessment activity according to criteria

, described below.

It is the audit team's judgment that Sequoyah Nuclear Power Plant meets the Supplement I to NUREG-0737 requirement for a comparison of display and control requirements with the control room inventory.

2.4 Control Room Survey The key to a successful control room survey is a systematic comparison of the control room to accepted human engineering guidelines and human factors principles. One accepted set of human engineering guidelines is j provided in Section 6 of NUREG-0700 (Reference 4); however, other accepted human factors standards may be chosen. Discrepancies should be documented I

as HEDs.

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The human engineering guidelines used for the control room surveys were a modified version of Section 6 of NUREG.0700. Modifications to the checklists were primarily alterations of general guidelines to make them plant specific. Clarifications of the guidelines were made as appropriate.

In addition, operator interview questions were referenced in the guidelines so that the person performing the survey was able to coordinate the operator interview questions and survey guidelines. It is the audit team's judgment that the survey guidelines and process for conducting the survey are comprehensive and thorough.

The survey activities were conducted from September, 1984 to February, 1986. During that time, there was a six-month break in order to transfer personnel resources to the Watts Bar Station DCRDR project. No deterioration of survey results were noted as a result of this break. The surveys were conducted in the main control room and the auxiliary control room.

In order to verify the comprehensiveness and accuracy of the survey results, the audit team conducted a sample survey in the Sequoyah Unit I control room. The object of the sample survey was to identify a broad range of typical survey checklist HEDs. The audit team identified 17 HEDs, categorized them by NUREG-0700 guideline number, and compared them to the DCRDR listings of HECs. As a result of this comparison, TVA was able to identify in their DCRDR HEC files all of the 17 HEDs identified during the sample audit. (See Attachment 6 for the results of the sample survey comparison to the Sequoyah DCRDR HEC files).

It is the audit team's judgment that TVA meets the Supplement 1 to NUREG-0737 requirement for a control room survey.

2.5 Assessment of Human Engineering Discrepancies (HEDs) to Determine Which Are Significant and Should Be Corrected Based on the guidance of NUREG-0700 and the requirements of Supplement I to NUREG-0737, all HEDs should be assessed for significance. The potential for operator error and the consequence of that error in terms of plant safety should be systematically considered in the assessment. Both the individual and aggregate effects of HEDs should be considered. The 8 1

1 result of the assessment process is a determination of which HEDs should te corrected because of their potential impact on plant safety. Decisions on whether HEDs are safety-significant should not be compromised by 1 consideration of sur.h issues as the means and potential costs of correcting HEDs.

l The Sequoyah DCRDR assessment was conducted by the DCRDR study team from March through August, 1986. Prior to beginning the assessment process, the DCRDR study team reviewed the assessment criteria, critiqued examples of l scaling HEC significance, and agreed on the methodology. The procedure for performing the assessment including criteria for determining HED priorities, (SQA179, Revision 1, Standard Practice, Appendix G) is provided as Attachment 7 to this report.

The first effort of the assessment team was to review, modify and document the assessment procedure. It was during this period that the category definitions were finalized. This document emerged as a useful tool I to evaluate HECs/HEDs in a standardized manner. The assessment process is described below. .

I Grouping of Related HECs - The assessment process began with the initial grouping of related HECs. Because of the way checklist identification numbers were organized, the initial grouping was accomplished automatically. For example, all HECs generated from the control checklist had the same numerical prefix,4,(e.g., 4005). Therefore, all HECs  !

generated by this checklist were grouped together. HECs were further 4 grouped on the basis of similarity or related concern.

Transfer From HEC to HED Status - After the assessment team members grouped the HECs, they determined which HECs qualified as HEDs. The criteria used by the assessment team to transfer the HECs to HED status are listed below:

1. The HEC must be related to the operation of the control panels in the main or auxiliary control rooms or transfer to the auxiliary control room. Control operation includes the environmental and communications functions.

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2. The HEC must - apply to the plant (i.e., not the simulater or

! mockup).

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3. The HEC must represent an actual deviation from a human factors criterion.

Some HECs were not assessed as HEDs. These HECs were assigned to one of two groups: Plant Concerns was a grouping for HECs not in the technical scope of the design review. Several concerns about procedures were placed in this group. These HECs have been forwarded to the site director with the recommendation that the HECs be evaluated by the appropriate supervisor (e.g., training, procedures). The supervisors are then to report back to the site director on the result of their review. Finally, the site director is to provide a copy of these reviews to the DCRDR manager, 'who reviews the resolutions for possible interactions with control room equipment.

The Nonacolicable Concerns group received HECs that were not valid.

Concerns that had been corrected since the HEC was submitted are an example l

( from this group.

HED Categorization - The assessment team applied a formal evaluation procedure and set of criteria to all HECs and HEDs resulting from the data gathering phases of the Design Review. They are dnscribed in the document, l "DCRDR Assessment Criteria and Guidelines" (SQA 179, Appendix G).

The categorization of an HED required several steps, which were recorded on an HED Categorization Worksheet (Attachment 7). First it was necessary to determine whether a postulated operator error resulting from the HED would have an effect on a critical safety function. Subsequent steps then examined (1) the likelihood that the error would occur, (2) the result of the uncorrected error if it occurs, and (3) the effect of the error on maintaining and/or restoring a critical safety function. Each of l these three evaluations was recorded on seven-point rating scales, the combination of which resulted in the categorization of the HED. To aid the team in the scale evaluations, the assessment procedure provided each scale  ;

J point on the three scales with a definition and example. The scales l provided an opportunity to record differences of opinion on the degree .of rating intensity. The resulting category was documented on the worksheet.

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l If a team member did not concur with the majority on the resulting  !

categorization of the HED, the dissent was recorded and explained on that j worksheet by the dissenter. The audit team reviewed sample cases of dissenter assessments and determined that the procedure was followed correctly. l l

In order to objectively determine which concerns should receive the most immediate attention for corrections, a set of prioritized categories were developed based on (1) reference to Sequoyah Technical Specification I limits and (2) applicability of the six Critical Safety Functions of )

Sequoyah Nuclear Plant. An HED could be assigned to one of four categories in which a Category I HED has the highest priority for concern and for correction. An additional criterion, that of impact on normal operations, was also included in Categories 3 and 4, as shown in the category definitions below

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l Category Definitions ,

l Category 1 l

Discrepancies in this Category relate to: 1) exceeding Technica' l Specification limits, which requires immediate action or 2j l

directly challenging or causing a loss of a critical safety function.

Category 2 Discrepancies in this Category relate to: 1) exceeding Technical Specification limits which allow limited time for corrective action or 2) reducing or causing the loss of resource (s) needed to maintain a CSF.

Category 3 Discrepancies in this Category relate to: 1) exceeding Technical Specification limits, which allow extended time for corrective action required or shutdown of the reactor, 2) adversely affecting normal operation, or 3) having the potential to affect a CSF.

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i Category 4 Discrepancies in this category have no significant effect on plant safety or operations.

l The definition of a safety significant HED encompassed the criteria of both HED categories 1 and 2. A safety significant HED was defined as any l HED which could result in a plant condition that (1) exceeds a technical l limit which required a corrective action within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or less or (2)

! challenges, reduces, or causes the loss of resource (s) needed to maintain a critical safety function.

During the assessment activity at Sequoyah, 1399 HECs identified by the DCRDR study team were reduced to 455 HEDs. The 455 HEDs were grouped into 10 Category I HEDs, 45 Category 2 HEDs, and 273 Category 3 HEDs and 12 7 Category 4 HEDs. Audit team review of all Category 1 and Category 2, HEDs and a sample set of Category 3 HEDs indicated that the assessment activity was appropri.ately conducted.

The audit team evaluated all category 1 and 2 HEDs in order to verify their safety significance (see Attachment 8 for the list of Category I and 2 l HEDs). In addition, the audit team evaluated a sample set of category 3 l HEDs in order to verify accuracy of TVA's assessment process.

i During the assegsment activity, 91 HECs were categorized c relating to training operations or emergency operating procedures (EOPs) rather than control room design. In order to ensure that these 91 HECs were resolved, the DCRDR team sent the HECs to site management for disposition to the training and E0P departments on November 26, 1986 (See Attachment 9). The response from the Sequoyah Division of Nuclear Training regarding the l training-related HECs, dated February 9,1987, is included as Attachment 10.

In addition to the assessment of the Sequoyah HECs, the DCRDR study team also reviewed all Watts Bar Nuclear Station DCRDR HECs to assess the applicability to Sequoyah station (see Attachment 11). The audit team I concluded that there was close coordination between the Watts Bar DCRDR results and Sequoyah results, and that applicable Watts Bar HECs were considered for SNPP.

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It is the audit team's judgment that TVA has conducted a thorough assessment activity that meets the Supplement 1 to NUREG-0737 requirement ,

for an assessment of human engineering discrepancies.

2.6 Selection of Design Improvements  !

The purpose of selecting design improvements is to determine I corrections to HEDs identified from the review phase of the DCRDR.

Selection of design improvements should include a systematic process for  ;

the development and comparison of alternative means of resolving HEDs. l Furthermore, according' to Supplement 1 to NUREG-0737, the licensee should l document all of the proposed control room changes. I The DCRDR assessment team developed and proposed a corrective action l for each HED in categories 1, 2, and 3. After completion of the corrective actions, each proposed modification was evaluated by TVA management. Of the <

328 proposed corrections, it was determined that 299 (91%) should receive corrective action. All category 1 and category 2 HEDs are to receive corrective action because of.their safety significance. Twenty nine non-  ;

safety significant HEDs will not receive corrective actions. I The selection of design improvements by the DCRDR study team focused on integrated corrective actions. The corrective actions fall into six general areas: .

l Control Panels Labels / Demarcations / Mimics Lighting Communications Annunciators Computers In many cases, Category 1, 2, and 3 HECs are being grouped into integrated corrective actions that will take place during the Cycle 4, 5, and 6 refueling outages.

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The DCRDR study team developed design modifications on a panel-by-panel basis. Full scale prints of the modified panels were generated by computer graphics. The prints included the revised panel layouts, l abel s, demarcations and mimics. In order to verify that they were making the appropriate changes, the DCRDR team used eleven Sequoyah reactor operators and other Sequoyah personnel to evaluate the adequacy of the proposed modifications.

At the time of the audit, TVA was still conducting studies on lighting, computers, communications, and annunciators. The schedule for completion of each of these studies is shown in Figure 1.

Based on Summary Report Figure 6, TVA indicated that the proposed DCRDR modifications would not be completed until 1992. This five year DCRDR implementation schedule was of concern to the audit team. In order to help resolve the audit team's concern regarding implementation of control room modifications, TVA provided NRC with Figure 1, CRDR Project Schedule.

In order to determine the adequacy of the proposed modifications and schedules for implementation, the audit team evaluated all Category 1 and 2 HEDs (Attachment 8) against the NRC guidance provided in Appendix A of NUREG-0800 Standard Review Plan Section 18.1. In addition, the audit team evaluated all Category 1 and 2 HEDs against the March 13, 1987, Sequoyah

" Restart Requirement Criteria" (see Attachment 12).

In most cases, the audit team agreed with the proposed HED correction and schedule for implementation. However, in a number of cases, TVA committed to modify the proposed correction or implementation schedule. The modified TVA commitments are provided on the following pages.

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j , MODIFIED HED CORRECTIVE ACTION i

l HED 0220 Narrow range containment-annulus PdIR needed inside horseshoe for resolution and present trend.

Correction:

Rodify correction to indicate that an interim correction of the containment annulus PdIR on M-9 includes:

1. Programmed release rate packages are now in place so the operator no longer has to wait for the chemistry f department to make the calculations. This is designed l to allow the operator to take timely corrective action, i

• and thus avoid a technical specification violation when containment reaches .3 Psi.

2. Increased surveillance of the current PdIR on the panel M-9 has been implemented.

3. Modified operator training related to previous licensee l event reports of technical specification violations and l the need to monitor the existing back panel meter will be implemented.

! By letter dated July 14, 1987 (Reference 9), TVA docketed a description HED-f 0220 interim corrective actions before mode 2 entry on unit 2 (see l Attachment 13). It is SAIC's judgment that TVA confirmed in writing the verbal commitments regarding this HED during the audit. is SAIC's

{ It 1 judgment that this is an acceptable interim corrective action.

HED 0379 Absent main steamline radiation monitors are needed, j . Correction:

The implementation schedule for this modification will be

! changed frer: refueling Cycle 5 to 4 to coincide with the Reg.

Guide 1.97 initiative. I l

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..y HED 1098 Control room emergency lighting levels are below the l HED.'1099 recommended level and normal lighting levels on 1-M-9 and 1-  !

M-10' does not rieet minimum (20 foot candle) lighting-

[,; s requirement. )

j -

x .(orrection: , ,

, 3* TVA Mill accelerate the lighting stuCy so that modifications

-resulting ffom the study can be implemented during refueling 4

r Cycie 4.'

N -g HEDi 30?5 ^ ~lJhere are many alar:ns' that require the operator to send for s

l information from 'outside the control room for determining )

%, . required corrective actions.

+ ' ,u ,

l Correction: ._

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.s t TVA committed to prioritize the section of the annunciator )

4 study related to this HED in order to verify that no Category  ;

k- 1 HED.t exist within this problem. Any Category 1 HEDs will j

{ be adoressed in the Cycle 4 refueling outage.

SCHEDULE .TVA committed to correct, to the maximum extent possible, l lower pridity Category 2 and 3 HEDs during refueling outages 4 and 5, rather than Cycles 5 and 6 respectively. TVA also

~ committed ti submit to NRC a revised schedule for

\'

modification implementation reflecting this commitment, once  ;

s it has been approvea by TVA management.

, s Bsed on the a Mit team evaluation of all Category 1 and 2 proposed I \

modifier;tions, along with review of a sample of. Category 3 modifications and i schedJ1 e for implementation, it is the abdit team's judgment that TVA has

, conducted an appropriate program for selection of design improvements. In order to meet this Supplerent I to NUREG-0737 requirement, it was necessary Tor TVA to submit two confirmatory documents. The first one regarding

^

corrective action commitments for HED 0220 was submitted by TVA on July 14, 1987. The second confirmatory document, Which requires TVA management s

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i l- j approval, will include the TVA commitments regarding HEDs 0379, 1098, 1099, and 3015, hnd should also contain schedule commitments.

2.7 Verification that Selected Design Improvements Will Provide the Necessary Correction A key criterion of DCRDR success is a consistent, coherent, and effective interface between the operator and the control room. One good way to satisfy that criterion is through iteration of the processes of selection of design improvements, verification that selected improvements will provide the necessary correction, and verification that the improvements will not introduce new HEDs. According to NUREG-0800, techniques for the verification process might include partial resurveys of mocked-up panels, '

applied experiments, engineering analyses, environmental surveys, and operator interviews. The consistency, coherence, and effectiveness of the entire operator-control room interface are important to operator performance. Thus, evaluation of both the changed and unchanged portions of the control room is necessary during the verification process.

Based upon expertise of the individuals, DCRDR Team members were .

assigned responsibility for proposing corrective actions for each of the HEDs. The proposals for corrective action were presented to the whole of the DCRDR Team for evaluation against two primary criteria:

o The correct,ive action should resolve the original concern o The correction should not result in new concerns A formal review and approval process equivalent to the assessment and categorization methodology was employed. Corrective actions resulting in panel arrangements were mocked up in an iterative process. Full-size computer generated modified panel layouts were then evaluated by operators and human factors specialists.

Upon completion of the iterative proposal process described above, HEDs enter the formal plant engineering change procedures of preparation, review, and implementation, which includes an additional human engineering review (EEB-EP 22.32 R0, Human Factors Engineering - Design Review).

18

It is the audit team's judgment that the licensee meets the Supplement I to NUREG-0737 requirements for verification that selected improvements will produce the necessary correction.

2.8 Verification that Selected Design Improvements Will Not Introduce New l HEDs.

As discussed in Section 2.7 above, the implementation of HED corrective

! actions at Sequoyah go through a formal plant engineering change procedure for preparation, review, and implementation, which includes, a human engineering review (EEB-EP 22.32 R0, Human Factors Engineering - Design Review). It is the audit team's judgment, that TVA has a process for verifying that selected design improvements do not introduce new HEDs, which meets the requirement of Supplement I to NUREG-0737.

2.9 Coordination of Control Room Improvements With Changes From Other Programs, such as the Safety Parameter Display System, Operator Training, Reg. Guide 1.97 Instrumentation, and Upgraded Emergency Operating Procedures Improvement of emergency response capability requires coordination of the DCRDR with other activities. Satisfaction of Reg. Guide 1.97 requirements and the addition of the Safety Parameter Display System (SPDS) necessitate modifications and additions to the control room. The modifications and additions should be specifically addressed by the DCRDR.

Exactly how the modifications are addressed depends on a number of factors including the relative timing of the various emergency response capability upgrades. Regardless of the means of coordination, the result should be integration of Reg. Guide 1.97 instrumentation and SPDS equipment into a consistent, coherent, and effective contr31 room interface with the operators.

Evidence of the coordination of the DCRDR effort and the upgraded procedures was discussed in Section 2.2 on task analysis. The first task analysis used the information needs derived from the E0P development activities to perform the "first" control room verification of availability 19

and suitability. It was noted that the project manager for-the DCRDR is the supervisor for the E0P development team.

The audit team reviewed those Human Engineering Concerns that were l

corrected by training methods and noted that all actions taken by the l training department were substantive. The audit team also noted that any plant design changes, such as will ensue from implementation of HED

' modifications, are reviewed and addressed as subject material for future training by the training department. ,

Sequoyah has assigned the lead engineer responsibility for the Reg.

Guide 1.97 modifications to the DCRDR team, thus ensuring that any ]

modifications due to Reg. Guide 1.97 requirements are also under the f cognizance of the DCRDR team. The audit team also noted that all control )

room panel modifications including Reg. Guide 1.97 instrur.entation fall l under the cognizance of the DCRDR team prior to implementation. I The design and implementation of SPDS-incorporated human factors guidelines was subjected to the review processes of the DCRDR methodology.

However, due to significant software and hardware problems experienced by the licensee with the SPDS since its implementation, a major effort is in progress at the present time to enhance reliability. To overcome basic I

design inadequacies, the licensee is considering the complete replacement of the current SPDS with an upgraded system. The project supervisor, for the 1 SPDS is working closely with the DCRDR team to assure coordination with the control room modifications. l It is the audit team's judgment that Sequoyah Nuclear Power Plant meets ,

the Supplement I to NUREG-0737 requirement for coordination of the DCRDR with other Supplement 1 improvement programs. 4

3.0 CONCLUSION

S TVA submitted the Detailed Control Room Design Review (DCRDR) Summary Report for Sequoyah Nuclear Power Plant Units 1 and 2, to NRC on November 26, 1986. A preliminary evaluation of the Summary Report was conducted by SAIC which resulted in the identification of a number of concerns. In order.

to resolve the concerns and evaluate the Sequoyah DCRDR, a pre-20

implementation audit was conducted from June 22 to June 25, 1987. During' the audit, the NRC staff, accompanied by SAIC and Comex representatives, ,

performed a detailed evaluation of TVA's DCRDR. The evaluation included examination of TVA's DCRDR' documentation, discussions with the DCRDR study l

team, inspection of the existing control room, and inspection of mockups and proposed corrective action modifications. This report reflects the consolidated findings and conclusions of the NRC audit team. The  !

conclusions are provided below, organized by the nine Supplement I to NUREG-l 0737 DCRDR requirements.

1. The establishment of the multidisciplinary review team used for the l DCRDR meets the requirement of Supplement I to NUREG-0737.

1

2. The system function and task analysis, which was based on Revisior,1 of the Westinghouse Emergency Response Guidelines and supplements, meets the requirements of Supplement 1 to NUREG-0737.

3. The control room inventory meets the requirements of Supplement 1 to NUREG-0737.

4. The control room survey methodology and results meets the requirement f

of Supplement 1 to NUREG-0737.

5. The methodology for and results of assessment of human engineering discrepancies meet the requirements of Supplement 1 to NUREG-0737.

6. It is the audit team's judgment that TVA has conducted an appropriate program for selection of design improvements. It is also the audit team's judgment that none of the DCRDR Category 1 or 2 human engineering discrepancies are significant enough to warrant delay in restart or limits on plant operation. However, in order for TVA to meet the Supplement 1 to NUREG-0737 requirement for selection of design improvements, it will be necessary for TVA to submit the confirmatory document described in Section 4.0 below, to NRC.

7. The methodology for verifying that contrd room improvements correct HEDs meets the requirements of Supplement 1 to NUREG-0737, 21

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8. The methodology for verifying that the control room modifications do 1 not introduce new HEDs meets the requirements of Supplement l' to ,I NUREG 0737.

9. The coordination of the DCRDR with other programs, including upgraded j E0Ps, SPDS, Reg. Guide 1.97, and training, meets the requirements of Supplement I to NUREG-0737.

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A list of DCRDR related Sequoyah employee concerns and the audit team's 1

evaluation of those concerns is provided as Attachment 14 to this Technical Evaluation Report.

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4.0 CONFIRMATORY DOCU"ENTATION NEEDED It will be necessary for the licensee to submit to the NRC the following documentation.

1. Report 1 - Submitted to NRC on July 14, 1987. Included as Attachment 13 to this Technical Evaluation Report.

2. Report 2 - Due when TVA Management approval received.

HED 0379 Absent main steamline radiation monitors are r.eeded.

Correction:

The implementation schedule for thir modification will be changed from refueling Cycle 5 to 4 to coincide with the Reg.

Guide 1.97 initiative.

HED 1098 Control room emergency lighting levels are below the HED 1099 recommended level and normal lighting levels on 1-M-9 and 1-M-10 does not meet minimum (20 foot candle) lighting requirement.

Correction:

TVA will accelerate the lighting study so that modifications resulting from the study can be implemented during refueling Cycle 4.

HED 3015 There are many alarms that require the operator to send for information from outside the control room for determining required corrective actioas.

Correctign:

TVA committed to prioritize the section of the annunciator study related to this HED in order to verify that no Category 1 HEDs exist within this problem.

23

SCHEDULE TVA committed to lower priority Category 2'and 3 HEDs during refueling outages 4 and 5. TVA also committed to submit a ,

revised schedule for modification implementation to NRC  !

reflecting this commitment, after it has been approved by TVA management.

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.0 REFERENCES

1. Letter from D.S. Kammer to E. Adensam, forwarding " Program Plan for Control Room Design Reviews for All TVA Nuclear Plants," June 9,1983.

2. Supplement I to NUREG-0737, " Requirements for Emergency Response Capability" (Generic Letter No. 82-33), December 17, 1982.

3. Letter from L.M. Mills to E. Adensam, forwarding TVA Program Pl an ,

September 13, 1983.

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4. NUREG-0700, " Guidelines for Control Room Design Reviews," September 1981.

5. NUREG-0801, " Evaluation Criteria for Detailed Control Room Design Reviews," Draft for Comment, October 1981.

6. Memorandum for: T. Novak, NRC, From: W. Russell, NRC,

Subject:

Review of Tennessee Valley Authority Program Plan for Control Room l Design Reviews, NRC, November 17, 1983.

7. Detailed Control Room Design Reviev Summary Report for the Sequoyah Nuclear Plant Units 1 and 2, Tennessee Valley. Authority Soddy Daisy, Tennessee, October 31, 1986.

I

8. NUREG-0800, " Standard Review Plan," Section 18.1, " Control Room," and Appendix A, " Evaluation Criteria for Detailed Control Room Design Reviews (DCRDR)," September 1984.

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9. Letter to: Document Control Desk, NRC, From: R. Gridley, TVA,

Subject:

Sequoyah Nuclear Plant (SQN) - Audit of Detailed Control Room Design Review (DCRDR) Program, TVA, July 14, 1987.

25 I

ATTACHMENT 1 LIST OF MEETING ATTENDEES

MEETING ATTENDEES NAME POSITION ORGANIZATION D. Bradley Engineering Specialist TVA/SQN J.F. Brooks Crdr. Imp 1. Manager TVA/DNE/ PES R.J. Burbbok Site Representative TVA/DNP M. Burzynski Reg. Lic. Manager TVA/ SON i J.M. Christens Sr. Electrical Engineer TVA/DNE/EEB/KN0X

! M. Cooper Licensing Engineer TVA/SQN J. DeBor Contractor SAIC/NRC J.J. Erpenbach Manager of Projects TVA/WBN T. Flippo Quality Surveillance Supervisor TVA/DNQA/SQN G.W. Gault Rx Engineering Supervisor TVA/SQN/TSSG C. Goodman CRDR Team Leader NRC/0SP E.F. Goodwin Tech. Assistant to Director NRC/0SP P.K. Guha Assistant Br. CH-EEB TVA/DNE/EEB P.E. Harmon Resident Inspector NRC T.L. Howard Acting Q.A. Manager TVA/SQN/DNQA G.B. Kirk Compliance Licensing Manager DNSL/SQN J.T. LaPoint Deputy Site Director TVA/SQN B.L. Mabry Instru. Engr. Comp. Eng. Section TVA/DNE/0ES/CES A. Marinos Chief-Reactor Operations NRC/0SP J.A. Martin Senior Human Factor Specialist TVA/EEB T.M. Nobles Plant Manager TVA/SQN A.M. Qualls Assistant to Plant Manager TVA/SQN R.R. Reeves Reg. Coordinator - EEB TVA/DNE/EEB R.N. Scheide Lic. Engineer TVA/SQN D.H. Schultz Contractor COMEX/SAIC/NRC R.E. Smith DEA TVA/DNE/EEB J.F. Tortora Principal Electrical Engineer TVA/DNE/EEB R. Vandergriff Operations TVA/SQN T.J. Voss H.F. Specialist Essex J.R. Walker Assistant Manager Operations TVA/SQN/ Operations

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ATTACHMENT 2-MEETING AGENDA 1

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I TENTATIVE AGENDA FOR SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 AND 2 DETAILED. CONTROL ROOM DESIGN REVIEW (DCRDR)

PRE-IMPLEMENTATION AUDIT JUNE 22-26, 1987 DAY 1 8:30 AM IntroductionandBriefing(NRC).

9:00 AM Overview discussion of DCRDR activities and results-(Licensee).

10:30 AM Brief tour of control room.

11:00 AM Requirement 1 - Establishinent of a qualified multidisciplinary review team.

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• Review of team management.

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• Review of teain members.

• Revienofteammember's(particularlyoperationsand human factors) roles in each DCRDR activity.

12:00 Lunch 1:00 PP, Requirement 2 - System function and task analysis.  !

• Review cf opgraded E0Ps used for tne task analysis (Rev. I to the Westinghouse Owner's Group Emergency

, Resporise Guidelines).

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• Determine if a comprehensive task analysis of all operator performed during emergency operations. This includes all Warning, Caution and Note tasks that are to b'e performed by the operators.

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• Audit a sample set of task analysis documentation.

The audit team will evaluate the complete set of documentation for procedure for

• Reactor Trip cr l Safety Injection."

l 3:00 PM Requirement 3 - Comparison of display and control require-ments with a control room inventory.

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• Determine if a comprehensive inventory of all control room instrumentation was conducted.-

• Review the results of the verification of the avail-ability of controls and displays to meet the operator's )

information and control needs. ,

• Review the resL}ts of the verification of the '

i suitability of controls and displays to meet the l operator's information and control needs.

5:00 PM End of review activities /NRC caucus.

1 DAY 2 8:30 AM Requirement 4 - Control room survey

• Review of the survey checklists to verify that they confccm to guidance such as NUREG-0700.

• Review cf the survey results.

10:00 AM Sample survey conducted in the control room (1 to 2 hrs.).

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• Comparison of sample survey conducted by NRC auditors to licensee's results.

I 12:00 Lunch 1:00 PM Requirement 5 - Assessment of human engineering discrepancies (HEDs) to determine which are significant and should be corrected.

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• Review the assessment process.

• Evaluate how consistently the assessment process was i

applied to category 1 and 2, safety significant HEDs.

• Determine how the cumulative effects of HEDs was identified and assessed.

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1 3:00 PM Requirement 6 - Selection of design improvements.

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• Review the selection of design improvement process.

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• Evaluate proposed control room design modifications.

• Evaluate proposed enhancement modifications.  ;

• Evaluate proposed training modifications used to correct HEDs.

• Evaluate proposed training modifications used to correct l

HEDS.

1 5:00 PM Break /NRC caucus.

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DAY 3 l

8:30 AM Requirement 6 (continued) - Review the selection of design improvements for all safety significant HEDs.

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• *10 Category 1 HEDs.

• 46 Category 2 HEDs.

11:30 AM Evaluate the implementation schedules for control room l modifications.

12:00 Lunch 1:00 PM Requirement 7 - Verification that selected design improvements ccrrect HEDs.

• Review and assess the licensee's process for varHying that the proposed modifications correct the HEDs.

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l-l Requirement 8 - Verification that the selected design 1:45 PM improvements do not introduce new HEDs.

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• Review and assess the licensee's process for verifying that the proposed modifications do not introduce new HEDs.

i 3:00 PM Requirement 9 - Coordination of DCRDR with other Supplement 1- l to NUREG-0737. initiatives.

• Review coordination with upgraded E0Ps.

• Review coordination with SPDS.

• Review coordination with operator training.

• Review coordination with Reg. Guide 1.97 requirements'.

5:00 PM Break /NRC caucus.

DAY 4 8:30 AM Operator interviews Operator interviews are conducted to determine how well the ;

control room design review addressed the operator needs and

. concerns. Interviews last from 30 minutes to 45 minutes.

1. Interview a shift supervisor

2. Interview a senior reactor operator

3. Interview a shift technical advisor  !

12:00 unch 1:00 NRC Caucus

• Determine final conclusions regarding the nine Supplement 1 to NUREG-0737 Requirements.

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• Determine the final conclusions regarding the operators' Concerns.

• Determine the final conclusions regarding the allegations. J 4:00 PM NRC audit team /TVA DCRDR team technical discussion. l

• Resolve open issues.

• Verify that' TVA personnel are aware of all technical concerns and what it will take to resolve those concerns.

1 5:00 PM Break / prepare exit briefing

,0?Y 5 9:00 AM Exit Briefing i

• NRC position with regard to nine Supplement 1 to NUREG-0737 Requirements at Sequoyah  ;

• NRC position on DCRDR related concerns

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• NRC position on DCRDR related allegations t

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A Standard Practice Page 360 Appendix G 89^1 9 l Revasion 1 1 III. Dgtailed Di?cussion_ of Assessment criteria (contanued) ..

The fif th part is verification and validation of each modification proposed as a result of a HED. Proposed modifications are made on

. mockups or drawings and evaluated with procedures (e.g., emergency i operating instructions, system operating instructions) to determine their overall effectiveness. Experienced operators are used to evaluate proposed modifications or alternat' , approaches. Established human factors guidelines are also used to r6eisw proposed modifications.

IV. CRSR Assesswent Work Session Procedure ]

Phase I - Grouping of Concerne The first phase is partially done as the HECs are written. The HICs {

from the same checklist are already grouped together and the HEC number I gives the panel unless the HDC applies to more than one panel. The HECs from task analysis are also gr'ouped together.  ;

Pha s e II - HE0 to HED The HICs being assessed wall be first evaluated ac to validity. If a l HEC is a valid concern and within the scope of the CRER effort, it will be either upgraded into a HED or incorporated into a HED with other related HID(s).

The following are criteria for determination of HIC to HID status (excep: as noted above):

1. The HEC must be related to the operation of the control panels in the rain or auxsliary control rooms or transfer to the auxiliary l i

control room, and

2. The HEC must apply to the plant (i.e., not the simulator or mockup), i i

3. The Hrc must represent an actual deviation from a human factors I i

craterion.

Tor assessnent purposes HED groupings by review area will be used (e.g.. displays, con unications, panel layout, task analys:s).

However, data base sorts by area or system will also be available to help ensure HED interrelationships are considered. Many HIDs may be identical or closely related. . Actions will be taken to combine similar .

HIDs into one HID. Thas will also facilitate consideration of HED interraJationships and provide for similar solutaons to related HEDs.

A combined HED nurber will be selected from one of the HEDs being combine d, The number will he the one that best describes the concern.

These assignments will be made by the CRDR team leader or designated team menbers. Majority to'am approval is required for combining HIDs.

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J Standard Practice Page 361 Appendix G 3@A17' Revasion 1 Phase III - Assessrent Worksheet Thie phase involves *ka completion of the asses--en', worksheet. The j team will complete the assessment worksheet.

The KED assessment that will be obtained will rate the HED with a priority (i.e., into one of four categories). The higher the priority, the greater the potential impact of the HED on plant operations. This will be described in detail below. i i

If other HIDs relate to the HED being assessed in such a manner to make l the KED more likely to occur, then it will be so noted on the HID. I Phase IV - HID Correction This phase will be accomplished by assignments made to team member (s).

The assigr.ments will be rade to individual team members or a group i depending upon the scope of affort ' involved. The correct 2ve action l will then be reviewed by the entire team for agreement. Each HID (i.e., Category 1, 2, & 3 HEDs) will be providtd a recommended corrective action. The corrective action will be entered on the HID form (Attachment 4). The team will sign off and date the HID form.

Phase V - Review of Corrections After completing the corrective action recommendation the team will use a ec:kup or a plant specific aimulator to review the changes re:or. ended. The fo:us of this review would be the corrective action selvan; the original concern and the possibility of other HICs being created.

The team will then request that plant operators review the corrective actions recernended. These operators should not have been involved in prev 2eus CRDR work except for questionnaires and ir.terviewc. Operator co.ments will be utiliced to finalize the corrective acticn.

I V. CRDR Work Sessiens ]

)

A ma]ority of the CRDR members (or an alternate that is very familiar I with what the team wall be meeting to discuss) most be present to have a quorum. The absent merter must review all HICs as potential MID(s) and subsequent HID corrective action determined assessed during his absence. If the absent team member does not agree with actions taken at a meeting that was massed, he must neke his disagreement known in ,

writing,as soon as possible and sign a,11 assessment worksheets. {

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Standard Practics Page 362 )

Appendix G 89A179 Revasion 1 VI. Procedure for Completion of Assessment Worksheet -,

A. purpose of Worksheet The worksheet ( Attachmer..1) isdesignedtoprovideamethodof determining each HID(s) potential for causing or contributing to l operating crew error and the consequence of such error on plant safety and operation. This process culm: ites in the assignment of the HID to a estegery prioritized on a C .tical Safety Tunction ba sis. (see Table I).

I B. Detailed Discussion of Worksheet

1. The first step in completion of the worksheet is to decide (yes)

(no) if the HID has the potential to impact plant safety. Impact i on plant safety includes whether the MID could lead to a ,

violation of Technical Specification Limits, Operatsng Limits, or could adversly effect a Criti' cal Safety Function (CSF). The six . (

critical safety functions are identified in the Westinghouse Owner's Group Dmergency Guadelines, Revision 1, which are the basis of Sequcyah Nuclear stations emergency instructions. The following guidelines are used to determine potential to impact plant safety: .

a. The HID could either directly or indirectly lead to a violation of Technical Specification Limits, Operating )

Limits, or could jeopardize a CST. At this point. if it is decided that the HID could lead to a violation of Technical Specifications Limits or Operating Limits, then the HID will be upgraded to a Category 1, 2, or 3. The team will decide the spec Fie category (1, 2, or 3) based upon expert judgement concerning the likelahood of error and seriousness of the error as defined by the Technical Specifications (see Attachment 6).

b. Whether or not the MID would have any impact on the sax barraers to radiataen release: saberiticality, core cooling, heat sink, pressurized thermal shock, containment intregrity, invento ry.

2. The worksheet ner.t rates the HID for its likelihood for causing an error. The team will rate the likelihood the HID will cause an error. The various points on the scale are defined in -

Attachment 2.

3. The next item for the team to rate is the result of an error if the, error is uncorrected. Attachment 2 gives the definition of the various scale divisions. Attachment 3 iists factors which are considered an th*e final categorization of the HED in regard to this item.

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Standard Practice Page 363 Appendix G SQA179 Revasaon i VI. Procedure __for Completion of Assessment Worksheet (continued) ..

4. The third step is the effect the error may have on the plant safety fanct4o.... Inese functions are delaned by the EOPs. The j .

scale mark definition is provided in Attachment 2.

$. The worksheet will contain the

• rating on each scale of each team eeA3er's rating. The decisisn as to wSet category the HID fal.'r into is by simple majority of the team member ratings. Each team member will be assigned a lettar that will be placed on each scale. Each team member will verify the correctness when sign of f is made on each HID. This will be done at the time of the assessment (i.e4, the assessment worksheets will not be typed).

HED(s) that fall within the CAT-4 part of the first scale are not evaluated f 2rther unless thay have a potential to impact a CSF.

The same occurs for HED(s) that fall within CAT-3 part of the second scale. ~

6. The CRDR team nenbers then each sign the assessment worksheet indicating their concurrence on the assessment done in steps 1 through 6. If any team eerber does not agree, it is so noted with a brief form. This may (1 or 2 sentence) reason on the HED assessment

- be followed by the dissenter preparing a detailed written response to the team leader. This written response must detail exactly what the team center is not agreeing with and any reasoning that eay help a third party understand the dissent.

This process is shovn in Attachment 5.

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Standard Practice Page 364 Appendix G SQA179 Revision 1 Table I ..

i Category Definitions " I Category 1 Discrepancies in this Category rt .e s : 1) to exceeding Technical Specification limits, which requires imeediate action or 2) Directly challenge or cause a loss of a CST.

Category 2 Discrepancies in this Category relates: 1) To exceeding Technical Specification limits which allow" limited tire for corrertive action or 2) Reduces or causes the loss of resource (s) needed to maantaan a CST.

Category 3 Discrepancies in this Category relates: 1) To exceeding Technical Specificat2on limits, which allow extended time fer corrective actien required or shutdown of the reector or 2) adversly affects normal operation or 3) have the potential to affect a CST.

Category 4  ;

Discrepar.caes in th2s category have no significar.t affect on plant safety or operat2ons.

Note: 1.) Safety Significant HEDs are found in Categories 1 and 2.

2.) Eere Categor2es 3 or 4 HEOs which involve sarple easy to do improvements may be upgraded to CAT-1 or -2 by CROR team vote.

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Attachment 1 Page 365 Appendix G SQA179 SEQUOYAH NUCLEAR PLANT Revision 1 I _ . . _ ,

CONTROL ROOM DESIGN REVIEW HED o HED CATEGORIZATION L __....

WORKSHEET C AT E G O.~4 DOES THE HED HAVE POTENTIAL TO IMPACT A CSF7 (Y/N)

LIKEllHOOD THAT HED WILL CAUSE ERROR ,

I CAT.4

, , , , CAT.1.,3,OR3 ,

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OEFINITELY VERY- 1 UNLIK E LY WAYBE LIKELY VERY NOT UNLIK E LY I LIK EL Y OEFinitTdL / l

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RESULT OF ERROR (IF UNCORRECTED)

CAT.3 l CAT.1,S OR3 L e n i e _ n a J s I a 8 '

a g g NO R E Q UIR E S REDUCTION LOSS OF l LOSS OF E,XTENDEO EXTENDEL EFFECT A D DITIO N A L IN OPER. 1 COMPONENT SYSTEM LOSS OF LOSS OF STEPS PERFOBWANCE FUNCTION FUNCTION SYSTEW~ PLANT FUNCTION FUNCTION

-.4 EFFECT ON MAINTAINING AND/OR RESTORING A CSF- I CAT.3 i CAT.2 l CAT.1 L e n a  ! A i a a i i 3 i i e NO POT E NTI AL REDUCEO LOSS OF g CHALLENOE LOSS l P R E V E r ' .* :

, EFFECT R E D U CTION l CSF CSF l TO A OF RESTORATi; TO CSF W AIN TEN A NC E W A INT E N A N C E CSF CSF W AINTEN ANCE RESOURCE RESOURCE RESOURCE C AP ABILIT Y  !;

REM ARK S/JUSTIFIC ATION: . ._ . . .

TEAiU ACTION '

C ATE GORIZ ATION 1 2 3 4 j NOTE: CISSENTING TE AW WEWSER(S) OPINf 0N NOTED ON SACK l TEAW WEMBER ,

-1 TE AM WEMBER $10N ATURE CONCURRENCE OATE JIM M ARTIN (L) YES NO G ARY O AULT (N) YES NO HFS (H) YE8 NO R AND ALL MclN10SH (I) YES NO RON VANDERGALFF (0) YES NO YE8 NO

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1 Page 366  !

l Attaereent 1 (continued) SQA179 l l Appendix G Revision 1 )

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i DISSENTING OPINIONS (COMMENTS): ,

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ATTACHMENT 8 CATEGORY 1 AND 2 HED CORRECTIVE ACTIONS l

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l HED PROPOSED CORRECTIVE ACTIONS i

CATEGORY 1 l HED 0220 Narrow range containment annulus PdlR needed inside horse-shoe for resolution and present trending.

Correction:

Retain PDl-30-133 on M-9 as constructed; add PdlR on M-6 to provide narrow range indication inside horseshoe and present i trend between normal operating limits. Also evaluate resolu-tion requirements.

This HED is considered to have a partial correction because one of the four recommendations from the proposed corrections which was to evaluate the visual resolution requirement was deemed unnecessary. Containment pressure indicator scales do not require a resolution capable of reading values of 1.54 psi and 2.81 psi because the SQN EOPs do not instruct the operator.

to take action at these exact values. Instead, the EOPs instruct s the operator to take action relative to such values, such as less than, greater than, and so forth.

HED 0301 The operator is not given direct indication of maximum verit time.

Cerrection: '

Program the TSC computer to perform this calculation if possible (depends on HED 8138 and HED 0312); revise the calculation sheet (Appendix A of FR-1,3) to specify where the recuired data can be obtained, and to include signoffs for the calculation performer and the second party verifier.

HED 0360 All containment isolation indication is not included on the status light panels.

Correction:

Provide complete containment isolation status on M-6 with status lights located in functional groups.

HED 1096 The Appendix R lights in the MCR shine into operator's eyes.

Correction:

The team recommends that the lights be redirected and that this system be evaluated to provide a design integrated with other MCR lighting.

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l , HED 1100 The control room standby lighting levels are below recom-mended levels (10 fc).

l Correction:

Redesign the standby lighting layout to provide an even distri- >

bution of adequate illumination levels and provide diversifica- 'g' tion of power sources. >

HED 2006 Co m m unica t ions-t elephen e-dif ficult to use phones while wearing protective gear (i.e., face masks).

Correction: _

l Replace present breathing mask with one whbh permits improved telephone and radio commun cations, HED 6001 The temporary labels, various number'ng, and setpoints on meters need to be made permanent.

Correction:

Provide permanent labels and setpoints on raeters and imple-ment a procedure for their control.

HED 6002 The nameplates are not the standard color and are difficult to read.

Correction:

k Replace non standard nameplates with those that are tbe standard white background with black letters.

l HED 6003 Specific labeling (nameplate) concerns.

Correction:

Replace current labeling with a uniform, standardized scheme l for labeling that incorporates standardized names, abbrevi-ations, acronyms and symbols.

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HED 8083 Emergency core cooling system controls are notgrouped well relative to operating sequence. <^

Correction:

Relocate ECCS components into clearly demarcated, labeled, functional groupings that adhere to 'A-B' ordering convention T  !

within groups.

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HED PROPOSED CORRECTIVE ACTIONS CATEGORY 2 HED 0021 Possible confusion of units during control room abandonment.

Correction:

Place small plaques on the end of the Reactor MOV boards on elevation 749 that follow the color coding for "A" train and "B" train boards.

HED 0210 The main feedwater bypass valves do not have control room indication of their status.

l Correction:

Add open/close lim.! switches to the FW bypass valves, and add status lights to M-3 which utilize a 2 x 4 matrix of MSC 800's with push to test circuits. Alternative

provide feedback of actual valve position.

This HED is considered to have a partial correction because clarification as to how SQN management desired to implement the alternative recommendations was needed. The recommended alternative to provide feedback of actual valve position was approved.

HED 0217 Layout, demarcation, grouping and order discrepancies on 0-M-27B.

Correction:

Relocate ::omponents into clearly demarcated, functional groupings, utilizing mimics, summary labels and eliminating mirror imaging (see proposed layout for 0-M-27B).

HED P219 Layout of M-27A is confusing due to mirror imaging, disasso-cintion of controls and the presence of "depowered" hand switches.

Correction:

Relocate components into a process flowpath mimic which eliminates mirror imaging and utilizes concise summary labels (see the proposed layout for 0-M-27A).

l HED 0221 The containment spray pump and valve handswitches do not follow A-B convention and are poorly arranged.

Correction:

Relocate suction valves under pump handswitches and relocate indicators above related controls, complying with 'A-B' ordering convention (see the proposed 1kyout for M-6).

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HED 0230 Indicativa of pressurizer spray valve position required.

Correction:

Provide the required indication.

This EED is considered to have a partial correction because clarificadon reptding the positioning of the spray valves needed. The r(commended correction was actually approved; indicator lights will be located per HED 8058 and the M-4 layout.

HED 0235 The RCS pressure indientors do not have the required range and resolution.

Correction:

Upon implementation of PAM ECN L6185, RCS wide range pressure will be displayed on demand in a digital format, Train A and Train B plasma display units are to be added on M-4.

HED 024B Numbering mismatch between HS-46-56A and FCV-1-51.

Correction:

Re-tag 55-46-56A and HS-46-56B to reflect association with FCV-1-51 as felicws:

HS-46-56A becomes HS-4611 A and HS-46-56B becomes HS 51 8.

HED 0292 The CRD fans and dampers violate the A-B-C-D order of convention.

Correction:

Reorder CRD handswitches into logical, demarcated and labeled groupings with A-B-C-D order within groupir.gs.

'lais HED is considered to have a partial correction because it addresses changes within the proposed re-layout for panel M-9.

The proposed re-layout for this panel shows some meters to be above recommended height guidelines. To ensure readability of meters, it was recommended that affected meters be tilted downward 150 with a tilt bezel. However, this re. commendation to tilt meters was not approved because the addition of PDR-30-133 to panel M-6 ' 'll resolve the problem of possible inaccuracy in reading .~ iters on Panel M-9. (This HED is related to HED B154).

1 HED 0303 A pressurizer relief ter.k level recorder is needed to determine trend as required in EOI.

Correction:

Add a level recorder on Panel M-4 or M-5 for pressurizer relief tank level, g modify TSC to trend this parameter, if the TSC's reliability is improved.

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HED 0320 Control room communications (horseshoe area).

Correction:

Install vertically mounted telephones on the control panels as follows: three full-function on the horseshoe panel, and at least two full-function phones on the common vertleal panels.

This HED is considered to have a partial correction because the proposed corrective actions are contingent upon the additional in-depth study of the communication system recommended by the assessment team in the report "A Preliminary Evaluation of l the Communication System at the Sequoyah Nuclear Plant" (July 1986). The detailed design study will allow corrective actions for the communication system to be impismented on an integrated basis. j I

1 HED 0326 RCS cold leg temperature must be manually recorded at hourly j intervals.

Correction:

Modify TSC computer to calculate heatup and cooldown rates-and make values available to operator upon request.

I HED 0327 RVLS indicators on M difficult to use.

Correction:

Implement one of the following options:

1) Install a microprocessor which will determine which indicator is appropriate for given conditions, and install an alarm for  :

Inadequate Core Cooling; j

2) Add hierarchical and descriptive labeling that will provide l l information as to when each meter can be utilized, including i expected indicated levels for different conditions and levels ,  ;

j and conditions defined as Inadequate Core Cooling.

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i HED 0379 Main Steamline rad monitors are needed.  !

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Correction: i Modify scope of ECN L6539 to include installation of equipment i

! that detects steam generator tube leakage. I HED 1005 Vacuum cleaner noise on midnight shif t.

Correction:

To minimize present noise levels, provide a quieter vacuum and permit only one vacuum cleaner to operate in MCR at any giveh time.

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HED 1022 Operator protective equipment needs. I Correction: l Have Health Physics (through instruction in IP-17) ensure that a suffielent supply of coveralls to fit all sizes of operators be mainteined in the HP supply cabinets.

HED 1023 MCR high noise sources.

Correction:

The six HECs identified four specific noise sources which are corrected through other design change activities. No further i correction is required.

9 HED 1039 Cramped aisles.

_ Correction:

Provide new Shift Engineer desk located to side of existing ECB desk with sufficient desk and seating area and adequate floor area to facilitate traffic flow.

HED 1098 The control room emergency lighting levels are below the recommended level.

Correction:

1) Increase lamp size or add light fixtures as necessary to  !

111uminate key panels to a level of 3 fe.

2) Relocate light fixtures above 2-M-1, 2-M-2, 2-M-3,1-M-5 and 1-M-6 to better illuminate these panels.

3) Ensure diffuser panels are kept clean by periodic main-tenance.

4) Periodically inspect and replace burned out bulbs.

HED 1099 The normal lighting of 1-M-9 and 1-M-10 does not meet the 19inimum (20 fc) lighting requirement.

Correction:

1) Locate some light fixtures over 1-M-9.

2) Perform periodic maintenance of the lighting system to ensure adequate lighting levels are maintained by replacing burned out bulbs and cleaning the ceiling panels.

HED 2001 Communications paging systems.

Corrections:

For each concern below, the following corrective actions are recommended:

Paging systems cannot be heard / understood in, some areas (2001, 0283).

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r o Provide loudspeaker with specialized functions to improve intelligibility. Place them in appropriately designated areas. i Distortion in paging system / telephone system and paging sys-tem not compatible (2002,2003).  ;

o Provide intelligible and compatible paging and telephone i syste ms.

A fast and reliable method of contacting auxiliary operators is required (2013).

o Provide an efficient and *eliable personnel paging system.

This HED is considered to have a partial correction because the proposed corrective actions are contingent upon the further in-depth study of the communication system recommended by the assessment team in the report, "A Preliminary Evaluation of i the Communication System at the Sequoyah Nuclear Plant" l (July 1986). The detailed design study will allow corrective actions for the communication system to be implemented on an integrated basis. ,

HED 2004 Communications - telephones - difficult to understand con- (

versations from high noise areas.

Correctiom  ;

Provide noise shielding / cancelling apparatus to enhance com- l f munications in high noise areas.  !

. This HED is considered to have a partial correction because the ,

proposed corrective actions are contingent upon the additional j in-depth study of the communication system. recommended by the assessment team in the report "A Preliminary Evg.luation of J I

the Communication System at the Sequoyah Nuclear Plant" 1 (July 1986). The detailed design study will allow corrective actions for the communication system to be implemented on an integrated basis. 1 i

l HED 2008 Communications sound powered phone system-Improve system and equipment.

Correction:

For each concern below, the following corrective actions are '

recoramended:

Sound powered phone headsets are restrictive, heavy and cum-bersome (2008, 2018).

o Use lightweight headsets on the Maintenance Jack system with standard headsets as backup.

Inadequate storage for the sound powered headsets (2024).

o, Provide sufficient cord length and headset storage at appro-priate maintenance and calibration stations.

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I Communications requirements for auxiliary control room may not be fully satisfied (2031).

o Provide additional communications for auxiliary control room per Appendix R requirements.

This HED is considered to have a partial correction because the proposed corrective actions are contingent upon the additional in-depth study of the communication system recommended by the assessment team in the report "A Preliminary Evalua1 Ion of the Communication System at the Sequoyah Nuclear Plant" (July 1986). The detailed design study will allow corrective actions for the communication system to be implemented on an integrated basis.

HED 2010 Comraunications radios - do not provide complete coverage of plantsite.

Correction For each concern below, the following corrective actions are recommended:

o Radio coverage of several plant areas is incomple,Te. These areas include ERCW, DG, CCW Intake, Cooling Tower, Turbine Bldg. and Auxiliary Bldg. (2010,2014,2015).

I Add additional Radiax cable runs and more repeaters to reduce the number of dead spots. Resolve existing problems with repeaters.

o Dedicated radio channel for Operations (2011). Provide Operations with their own dedicated radio channel.

This HED is considered to have a partial correction because the proposed corrective actions are contingent upon the additional

!n-depth study of the communication system recommended by the assessment team in the report "A Preliminary Evaluation of the Communication Systern at the Sequoyah Nuclear Plant" (July 1986). The detailed design study will allow corrective actions for the communication system to be implemented on an integrated basis.

HED 3009 Annunciator System - Too many alarms during a transient or accident.

_ Correction The annunciator system will be further studied for overall needs.

This HED is considered to have a partial correction because it involves further study of the annunciator system for overall need of system replacement.

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HED 3010 Annunciator System need blackboard concept.

Correction Apply blackboard concept to MCR annunciators only at full power operating conditions. (Note: if nuisance alarms are corrected per HED 3238, this HED requires no further cor-rective action.)

HED 3015 There are many alarms that require the operator to send for information to determine required action.

l Correction l

Perform a further evaluation of annunciators that requir~e information from outside the control room for determining required actions.

HED 3071 No alarm signal on permissive pt :el when they change status.

Correction Add a tone to the lightbox, _O_R replace the light box with a Ris series 3100 similar to the one on panel 1-M-30 for PAM.

HED 3077 Administrative controls are not used to control annunciator horn loudness.

Correction 1mplement administrative controls to regulate the adjustment of volume controls on horns, OR. procure or design an auto-adjusting horn volume control.

This HED is considered to have a partial correction because the loudness adjustment is achieved by an operator using a proesdurc vs. an automatic adjustment control.

HED 3085 Annunciator ACK/ RESET / TEST switches are not located con-sistently from panel to panel.

Correction Reposition poorly located ACK/ RESET / TEST switches as deter-mined by team.

HED 3173 Need better annunciation for ?hase A, Phase B, and contain-ment vent and indication of reset.

Correction Provide consistent Phase A, Phase B and containment vent isolation / reset indication on a signal basis.

This HED is considered to have a partial correction because the recommendation for reset indication (addition of amber and blue light circuits) was determined to be unnecessary. The

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r clearing of an annunciator window will provide indirect' indication whenever a signal has been reset.

HED 3233 Fire header low pressure alarm needed.

Correction Add a low pressure alarm to the Hi Pressure Fire Protection system.

HED 3238 Nuisance alarms.

Correction ~

Perform further evaluation of nuisance alarms when both units are running at 100% power.

i HED 4002 Accidental activation of controls.

Correction Install new guardrail system that will prevent accidental act!-

vation of controls and still provide control accessibility.

HED 4037 Faulty switch operation.

r Correction ECN L66339 corrected this problem by removing the " reset" switch position from subject controls; no further corrective j action recommended.

HED 4054 Plastic shields missing.

Correction Devise another method of sticking the holders to the metal benchboard.  ;

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HED 4066 Hold to operate switches are not clearly distinguished from j normal handswitches.

Correction Identify hold to operate switches with an "H"in the upper right corner of the label (this applies to MOVs only).

HED 406P Controllers eperate in different directionL to open valves.

Correction 1 Add labels to front face of controllers in order to 1) associate controller demand output to final controlled device position or direction of travel, 2) identify manual control direction of

( rotation to open valve, and 3)ldentify direction of rotation to

\ increase setpoi. t on rotating scale setpoint controls.

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e HED 5004 Meter and recorder scales - do not follow guidelines; are confusing; do not reflect engineering units operators need.

Correction implement a plant procedure that provides a standard for indicator scales based on human engineering guidelines, and change specific hardwsre as defined.

This HED is considered to have a partial correction because it was determined to be necessary for the Operations section to 1 have final approval of each corrective action contained within i this HED. If the corrective action is not considered necessary by the Operations section, it will not be implemented. l

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HED 5020 Use of zone coding.  !

Correction Implement a plant standard for zone coding to be used on MCR and ACR meters; implement a plant procedure for the applica-tion of zone coding.

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h;D 5113 Indicators used to display the various 6.9 KV board voltage do not have sufficient resolution to identify a degraded voltage '

condition.

- Correction Replace existing W VX 252 indicators with Dixon Instruments Model C101 which provides a bargraph and four digit readout; these indientors should provide operators with required resolu-tion.

HED 8059 Controls for pressurizer relief tank are disassociated.

Correction Relocate pressurizer related controls in the corner of M-4/M-5 to provide improved association of PRT components, and to enhance the relationship between controls and indicators (see the proposed layout for M-4 and M-5).

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HED 80S8 Pressurize tail pipe temperature indicator and acoustic monitor not grouped together.

Correction Remove XX-68-363 from Panel 0-M27A and relocate to Panel M-4 (coordinate with "re-layout" activity on M-4.

HED 8117 Low pressure safety injection valves not functionally grouped together or with rest of system.

Correction s_ Relocate HS-63-172, -93, -94 together in a demarcated area l with the following functional labels over each: for -172 " HOT 11

< LEG INJECTION," for -93 COLD LEG INJECTION," for -94

" COLD LEG INJECTION"(see proposed layout for IM-6).

HED 8138 Upper and lower containment temperatures must be calculated by operator.

Correction Modify either the TSC or P-250 computer to provide on demand the upper and lower containment temperatures; also, modify computer to trend this parameter.

HED 8154 Status lights for various ventilation / alt conditioning functional subgroups not associated with related controls.

Correction Relocate status lights from XX-5-9A, 9B, 9C to 1) handswitch module that controls damper 2) a functional location in a process flow mimic, and 3) adjacent to an associated control.

This HED is considered to have a partial correction because it addresses changes within the proposed re-layout for panel M-9.

The proposed re-layout for this panel shows some meters to be above recommended height guidelines. To ensure readab:lity of meters, it was recommended that affected meters be tilted

, downward 150 with a tilt bezel. However, this recommendation to tilt meters was not approved because it was determined that

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these meters do not require such accuracy in readings to justify tilting. (This HED is related to HED 0292)

(END OF CATEGORY 2 HEDs) s 12

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l ATTACHMENT 9 NOVEMBER 26, 1986 SQN-EOP/CRDR MEETING i

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.?V A 60 (b$.945) (OP WP4f" l'NITED STATES GOVERNMENT Memorandum TEWESSEE VALLEY AUTHORITY 70  : CRDR Files FROM  : J. A. Martin ..

DATE  : November 26, 1986 1

SUBJECT:

SQN-EOP/CRDR MEETING The purpose of this meeting was to formally discuss the CRDR HEC's that related to the SQN-EOP's.

The CRDR team performed a review on the 1399 HEC's found during the CRDR data collection phase. This review found that 91 HEC's were outside the scope of CRDR in that they were related to procedures, operation, or training. The CRDR team sent these HEC's to the site management for disposition.

These HEC's are found in attachment 1. The HEC's that were associated with the procedure makeup a large number of the HEC's.found in attachment 1. l l

These HEC's were felt to relate to or require a procedure change to resolve the HEC. Specific examples yere discussed. I

1. HEC 0271 is an example that implies a resolution problem with the meters, since the procedure uses a four digit No. 1117 psig, to be verified. The CRDR team by including this HEC in the out of scope HEC's does not feel that a instrument change is required to correct this concern. The CRDR team does feel that a minor clarification in the procedures would fully correct this concern.

2. HEC 0254 is an example that relates to nomenclature and on labels used in the procedures. HEC's relating to nomenclature in this group of HEC's were felt to require procedure change, not label changes.

The HECs in Attachment I with a "P" are procedure related.

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TVA MEETING PRESENTATION SLIDES l

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MAIN CONTROL ROOM AL JX ILI ARY CONTROL ROOM TASK ANALYSIB

. TVA TASK ANALYSIS

. ESSEX CORP TASK ANALYSIS 1

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ASSESSMENT RESULTE  !

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1.3 9 9 HEC' S REDUCED TO 455 HED' S 4

SUMMARY

i CATEGORY HED' S 1 10 8- 45 l 3 273 4 l- 127 we t

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CORRECTIVE ACTIONS

- CONTROL PANELS 4

- LABELS / DEMARCATION / MIMICS 1

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- LIGHTING  ;

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- COMMUNICATIONS

- ANNUNCIATORS O-

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-- I DCRDR STUDY TEAM PROGRAM MANAGER J. R. WALKER s

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TEAM LEADER ..

l J.A. MARTIN i NUCLEAR ENGINEER G.W. GAULT l

1 I&C ENGINEER C. R. McINTOSH ,

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OPERATOR,7 l l

l R.E. VANDERGRIFF T.J. VOSS CESSEX CORP) l l

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i DCRDR IPCEMENTATION TEAM  ;

PROGRAM MANAGER J.F. BROOKS 1

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1 R.E. VANDERGRIFF

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HUMAN FACTORS SPECIALIST l I

J.A. MARTIN

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ATTACHMENT 4 PROCEDURES EVALUATED DURING TASK ANALYSIS l \

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- SEQUOYAH NUCLEAR FLANT EMERGENCY INSTRUCTION INDEI

, s Instruction Revision Title

.i E-0 1 Reactor Trip or Safety Injection ES-0.1 0 Reactor Trip Response  ;

ES-0.2 2 SI Termination ES-0.3 1 Natural Circulation Cooldown J E-1 2 Loss of Reactor or Secondary Coolant 'l ES-1.1 0 Post Loca Ceoldown 1

ES-1.2 3 Transfer to RHR Containinent Sump I ES-1.3 0 Transfer to Hot Les Recirculation E-2 2 Faulted Steam Generator Isolation E-3 2 Steam Generator Tube Rupture j ES-3.1 0 SI Termination Following SGTR ES-3.2 1 Post - SGTR Cooldown Using Back. fill ES-3.3 1 Post - SGTR Cooldown by Ruptured S/G Depressurization E-FOP 0 Foldout Page r,e -

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1. .45'f, 2

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SEQUOYAH NUCLEAR PLANT DERGENCY CONTINGENCY ACTIONS INDDC i .

i Inst ruction - Revision Title i

j

-# ECA-0. 0 0 $

Loss of All ac Power '

-#r ECA-0.1 0 Recovery From Loss of All ac Power Without SI Required

$ECA-0.2 0 Recovery From Loss of All ac Power With SI Required' l l

1 (4

i 1

l i

3. ".

Revised 08/21/85 .,. T ;l:?

a-J .

0484L/vbo  ::: .

.,. 1 . _: -

e

,I e

4

SEQUOYAH NUCLEAR PLANT FUNCTION RESTORATION GUIDELINES INDEI ,

Instruction Revision Title FR-0 3 Status Trees

! FR-S.1 2 ,y Response to Nuclear Power

. . Generation /ATWS

- FR-S . 2 0 5 Response to Loss of Core Shutdown FR-C.1 1 Response to Inadequate Core Cooling FR-C.2 0 Response to Saturated Core Cooling FR-H.1 1 Response to Loss'of Secondary Heat Sink FR-H.2 0 Response to Steam Generator Overpressure FR-H.3 0 Response to Steam Generator High Level ,

FR-H.4 0 Response to Loss of Normal Steam ~

Release Capabilities FR-H.S 0 Response to Steam Generator Low Level FR-P.1 1 Response to Pressurized Thermal Shock FR-P.2 1 Response to , Cold Overpressure Condition FR-2.1 1 Response to High Containment Pressure FR-Z.2 0 Response to Containment Flooding FR-Z.3 0 Response to High Containment Radiation FR-I.1 _

0 ._ . Response to High Pressurizer Level FR-I.2 0 Response to Low Pressurizer Level

[FR-2.3 1 Response to Voids in Reactor Vessel  ;/~

2..: -

k.

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__..-__7_--,__-- ,y-- ._,

i ATTACHMENT 5 SAMPLE TASK ANALYSIS DOCUMENTATION

3 EOP 'oEVELOP ENT i , CCCDR TEAU activities I ACTtVITIES l l

l COMPLETE TA$K REVIEW OF

$YSTEM FUNCTION Y WORKSHEET l' +l l

OWNER $

j jf F GROUP jf ACTIVITIES BRIEF PRESENTATION DEVELOPiANALY$t$ OF OF PLANT OPERATION .

CONTROL ROOM OF EOP UNDER REVIEW {

OPER AT OR T ASKS s

l Y .

INPUTIOUTPUT REQUIRE- PERFORM WALK. TALK WENTS FOR OPERATOR THROUOH TASA ACENTIFIED 1

V MONITOR OPERATOR APPROPRIATE DEClllON + CAPABILITIES TO

• POINTS ARE FERFORM TASK '

DETERMIN ED I E. REACH LOCATION l

v h i

1. 1 REVIEW CONTROLS &

KEY INFORM ATION +

• I 4 D!$ PLAY! AGAINST .

NEED$ ARE DEFlNED g INFO REQUIREMENTS

$EE NOTE 1 ,

YES Y l MONITOR TRAFFIC

-> PATTERNS, COMM. + .

(s . EQUIPM ENT AVAILABLE TO NO ALT ERN AT E M EANS IDENTIFIE D REQUIREM ENT$. ETC. l' SUP PO RT IN EOP TASK $

l NO DOCUMENT COMMENTS YES AND DEVELOP MECs y

DESIGN CH ANGE SEE NOTE 2.

y

+

PROVIDE FEEDBACK TO ECP TE.AM ,

l 4

Y A$$ES$ NEDs DEVELOP I

VERIFICATION HEDs & DETERMINE CORRECTIVE ACTION Y

VALIDATION l

l NOTE 1; Information re ;uirements identified to support task: units, range, occuracy, re solution, etc.

NOTE 2: An HEC can be generated and provided to the DCRDR team for evolustion of the design change.

(

FIGURE 3. TASK ANALYSIS BLOCK DIAGRAM 16

TABLE 3 HUMAN FACTORS GUIDELINES A. Operator Interface Feature (s) Used to perform the Task are Located Where Needed.

1. Are all of the controls, displays, arid indicatort that are required to perform this task present in the control room?

2. Are the controls, displays, and indicators grouped in accordance vin. .he information and control requirements of this task?

3. Are related controls and dispicys within functional groups arranged in the same relative positions? For example:

A display E display C display A- control E control .C control

)

B. Operator Interface Feature (s) Used to Perform the Task are Easy to Locate / Identify and Interface With.

1. Are the controls, displays, and indicators labeled according to the requirements?

2. Can the displays and annunciators used in this task be read accurately from viewing position of the vperator?

Can displays be read while operating associated controls?

3. Do !he displays give the operator direct, readily usable information?

Parameter values with required precision?

- Range, band, limit shown if the operator needs to know in/out of range / band, above/below limit, etc.

Trend information when needed?

Rate of change information when needed?

4. If instrumentation or control capabilities are unavailable under certain plant conditions, are there siternative means provided for the operator to meet task -

needs?

5. Is the display obscured while operating its needed associat ad control?

6. Are there any controls, control positions, or displays Wich the operator knows do not rerve any function?

+

18

~

CHEET .._. OF

'T.w? - .

, . . . . , _ (AIRS) .

T l' ACTION-INFORMATION PEDUIREMENTS

SUMMARY

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PLANT: * -

ORIGINATORS REVIEWER:_ n.________

DATE:

DATE:

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I  ! UN I T S : .. ._ ____- _ __G.#./f .

EF  ! PRECISIONs,,,______ g ,,____

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SYSTEM:  !

- ! N if T  ! COMPONENTS EF i RESPONSE TIME:

1. - ' ~ .,c 1- 'O-

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• 1 1 TYPE _C.__ ___________,___.

jp

! PARAMETER: TTLFLW r:: -. - . . ~ _ _ .__...

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% VERIFigATIQN ELIMMARY ELQ"M t t ACTUAL Rt I

REMARKS: J.

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1 TREN.

STATE / UNITS / RNG/

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..P G_ _ _ _C=T_ EF _ __ A" T . W R_k _E.1.RE C.T 1O_ N_ .VAWE_

M18 ..

GPM O-400 10 N B OBS > 390 NJ

• 2."- E-O 19 ( O "')

390 GPM O-400 10 >

E-O 25(25) OPS >

N/A 10 N A OPS > 390 GPM E-O 20(21) GPM O-400 10 Nl OE(OE1 B 0P5 > 390 N:

. E-1 > 390 GPM O-400 10 E-2 3C1(NA) A OBS N/A 10 N' 390 GPM 390-ES-1.1 18(1E) OBS >

400 390- 10 N DBS > 390 'GPM ES-1.1 2O(191 400 390- 10 NI 00 E DBS > 390 GPM '

FR-H.1 400 gj 50-150 50 t-D DBS = 50 GPM FF-F.1 01 GPM O-400 10 N 13(14) .A OBS N/A 390

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4_ 3.0 INPUT DOCUMENTATION l C.E ' '

.l.: .:.. -

Ils P, Westinghouse Owners Group Emergency Response Guidelines and Back-

[.k .

1.

ground Documentation, Revision 1. . ,

r(c'

e. .

2. Plant-specific Abbreviation List.

3. System Piping and Instrument Diagrams.

4. Final Safety Analysis Report.

5 Westinghouse Owners Group System Review and Task Analysis Documen-

a. '

tation.

g .

6. Other plant-specific documentation, as appropriate. ,

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PARAn 31tI:T10N STATEl W:ISI 2.T. REOPREC E6! TREE A01 'iEtt SYS to '06 c D'J1Pe 41 0. ~ ._

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h. ' E ACTION-INFORMATION REQUIREMENTS

SUMMARY

E ORIGINATOR: --.. .... DATE:

REVIEWER: __. .._4....

DA4E .... ___ l

2. . PLANT:

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REMARKS:

.--__ 1D _U2_._____1. .E6 EEL-_

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4. c. EYSTEM FthCT10N AND TASK ANALYSIS

.i.T . .

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51*.".* . L '

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9.:'.K: -

$ $" BEHAVIORAL ELEMENT VERB LIST

$f. l isi . . .

• Definition ,

72.*

I

- r Verb Application

.r Lw

.:a,. To attend visually to the presence of or the Info. Req. status of an object, indication, or event.

Observe e'

.. f To examine visually information which is l

~ Info. Req. j Read presented symbolically. '

l To visus!!y keep track of an object, Info. Req. indication, or event over time.

Monitor l l

To quickly examine an information source to Info. Req. ,

obtain a general impression.

Scan To be aware of the presence or absence of a Info. Req.

'.: Detect visual stimulus. '

To manually or verbally initiate a simple or Cont. Req.

' Start complex function, event, or activity.

j To manually or verbally terminate a simple Cont. Req. l Stop " or complex function, event, or activity.

To manually or verbally initiate a simple or Cont. Req.  !

Open complex function, event, or activity which ultimately results in a plantvalgte, component or breaker, plant components (e.g.,

demper, etc.) to assume an open state.

To manually or verbally initiate a simple or Cont. Req. complex function, event, or activity which Close l ultimately results in a plant component or (e.g., valve, breaker, plant components damper, etc.) to assume a closed state.

To manually or verbally initiate a elmple or Cont. Reg. complex function, event or activity which Adjust ultimately results in a plant component or m

plent components, g a plant condition, status, or dynamic to change state.

.t .

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VALUE RATE . ...... .

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ATTACHMENT 6 AUDIT TEAM SAMPLE SURVEY HUMAN ENGINEERING DISCREPANCIES

1 i

i AUDIT TEAM SAMPLE SURVEY 6/23/87

1. Temporary demarcation tape around reactor first out annunciator panel (XA-55-4D) safety injection signal tiles on panel 1-M-4.

NUREG-0700 Guideline 6.6.5.1.a TVA HEC #0201

2. The vertical and horizontal axes of the annunciator panels on panel 1-M-4 are not labeled with alphanumerics for ready coordinate designation of a particular visual title.

NURGE-0700 Guideline 6.3.3.3.c(1) TVA HEC #3018 and #3172

3. The reactor first out panel (XA-55-4D) visual alarm tiles for safety injection signals are not functionally grouped within the annunciator panel.  ;

NUREG-0700 Guideline 6.3.3.3.b TVA HEC #6002

4. The intermediate range nuclear power meter on panel 1-M-4 is labeled as count rate, rather than current which it actually displays. Labels should describe the function of equipment, NUREG-0700 Guideline 6.6.3.1.a TVA HEC #6003 and #6041 i

5. Instrument blocks for intermediate range and source range instruments 1-HS-92-5006 on panel 1-M-4 have low contrast grey labels with dark grey lettering.

NUREG-0700 Guideline 6.5.1.3.c TVA HEC #6002

6. The delta flux recorders on panel 1-M-4 have no engineering units on the recorder scales or the chart paper.

NUREG-0700 Guideline 6.5.4.1.a TVA HEC #5018 and #5004

7. The reactor coolant system pressurizer spray valve position indication (open indication by light ON, layout reserved in loop 1 and loop 2).

NUREG-0700 Guideline 6.8.2.3 TVA HEC #0229 and #0230

8. Benchboard timer located on panel 1-M-4 for operator's use has no label.

All controls and displays should be clearly labeled.

l NUREG-0700 Guideline 6.6.1.1 TVA HEC #4006

9. Charging pump arrangements on panel 1-M-5 are layed out C AA BB. This is not a logical panel layout.

NUREG-0700 Guideline 6.8.2.2 TVA' HEC #8100

AUDIT TEAM SAMPLE SURVEY 6/23/87 (Continued) 1

10. Containment pressure gauges (1-PDI-30-42, 43, 44, 45) on panel 1-M-6 j have different numerical progressions on adjacent scales. Displays 1 which are to be compared should have compatible scales.

NUREG-0700 Guideline 6.5.1.5.d TVA HEC #5143 i

11. Layout of the letdown, volume control tank, and blend controllers are )

mixed up left to right (close open, open close, open close, and close open) on panel 1-M-6.

NUREG-0700 Guideline 6.8.2.4 TVA HEC #4068

12. Baron concentration - chemical and volume control system range (PPM) label on panel 1-M-6 is covered by star wheel selector switch (1-HS 94) when being controlled by the operator.

NUREG-0700 Guideline 6.6.2.4.b TVA HEC #6067

13. Labels for switche: such as letdown orifice on panel 1-M-6 are loose and rotate on the switch bezel. When the functional control position marks are moved, the operator is uncertain of switch position.

NUREG-0700 Guideline 6.6.3.8.a TVA HEC #4029

14. Label for upper head injection Fill-Charging Pump Receive Value on panel l 1-M-6 is incorrectly spelled UNICHARGING vs UHI CHARGING. 1 NUREG-0700 Guideline 6.6.3.2.f TVA HEC #6003

15. Panel 1-M-5 label RCS PRZR relief tank control valve really is the vent from tank.

NUREG-0700 Guideline 6.6.3.1.a TVA HEC #6041

16. The P-250 process computer display panel has numerous red (lit) Group Log Alarms. As part of a " dark" board annunciator panel concept, these tiles should not be illuminated under normal operation.

NUREG-0700 Guideline 6.3.3.2.e TVA HEC #7009

17. Indicator lamps have a high burnout rate of about two "handfulls" per _

day.

NUREG-0700 Guideline 6.5.3.1.a(3) TVA HEC #5027

l i

i l

1 1

4 l

\

I I

ATTACHMENT 7 ASSESSMENT PROCEDURES l

l l

l L_____________________________ . _ _ _ _ _ _ _ _ _ _ .

k Standard Practice Page 359 Appendix G SQA179 Revasion 1 1

)

I. Scooe These criteria are to provide the CRDR team with the method of 1 ostermining whether a HEC is moved to HED status and the method for categorizing HEDs. (see section 8.1.3 of CRDR program plan).

II. Purpose of hssessment

• The objectives of the assessment phase of the CRDR are as tcIlows: )

Determine validity of HECs.

• i Evaluate and document the significance of the HIDs identified in the I

earlier dsta collection phases of the CRDR.

J l

l -

Reconcend resolution for HEDs (i.e., changes to the control room. {

) procedures, operator training, or* any com.bination thereof to resolve i these HI?s).

l 1 III, retailed Discussion of Assessment Criteria l There are five rain phases of the CRDR assesse. tnt effort. In the first i phase, the HICs found in the earlier survey effort are grouped into I

ident2fiable areas (panel, system. etc.). This facilitates the review by the CRDR team. Particularly, it allows consideration of cumulative effects of HED(s).

l The second phase is the review by the CROR team of each HEC to determine its validity.

~his verifies that: (1) the concern which may have come from the plant's simulator is in fact valid for the plant, (2) the item is within the CRDR scope of ac+.ivity. This verification of the HEC enables the concern' to become a HED.

"he third phase is the establishment of a priority for correction of the HID.

The fourth phase is the determination of the preferred corrective action for each HED recommended for correction. Some MEDs may be corrected by connetic surface enhancement techniques, such as changing labels or reorganizing procedures, adding demarcation lines or mimic lines, etc. Correction of other HIDs may require more extensive -

measures. If it is determaned that the correction must involve movement, modification. or addition / deletion of controls and displays, then the*se corrections are evaluated with other alternatives and with consideration of how the corraction(s) wall impact the existing control room (consistency and compatibility), correction of other HIDs, plant j availability, operator tra'ining and performanet, and procedures.

0195S/LDW

i HUdh ENSINEER!h5 CCCERNS SEGUOYAH NUCLEAR PLANT 49/15166 PAGE HEDIICATI DIS RIPil0N 1 l 1 lifHERE FOUC IPANEL ! HIC I !SYS !RES%Ut!C i 1  !

+...+....................................................+I ..............+......+.. 1

.......+....... .............,1AS I

101 ...{

P1ii i IA!THER! ITHE  !$ OPERATORS A h!ED TO REDUCE AMOUNT C 70 DEVILOP MECRY AltS MUST REGUIREt MI CR12E TO MIMORJIEALL IMP 10038 I

!  ! 1 i

1 IMMEDIATE -

10-1l 1

I  !  ! I i g

1  ;

i 1 lALAH R!S*0NSE - THE OPERATORS ARE RESP'Od!!LE 101 FOR IMP 10039 I I

! !K C ulh5 THE IMMEDIATE RESPONSES TO ALL ALARMS I 16 !

l i I i i 1 -

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18 !$DME OF THE AICMR OPERAi!N5 INSTRU;f!ONS (A0!s) 00 !O! i 1 " IMP 10040 1 1 1CT ICLME A C09LETE LIST OF ALL AFFECTED ECU!PMINT, T- i 18 ll 1 l i  !

!FOR EZAMPLE A0! 25 1

i i 1 l I I i

{

i 101 PII 1 !ICCCENSI MACES I Wh TO ChiTHE MAct ORAICRMR OPIRATINS IMP 10042 l INSTRUCTIONS l

PUT E0Ps AC A0is I ON i  !

!61) l ICCMPUTER  ! I 1 I  !  !  !  ! {

10!

P!! !I iPOSSll;Y1AICER 0FERAi!N5 IREA:108 C0hiFOL MAKE ERROR INSTRU:TICN SYSTEM' 15 EASY 70 6ET LOST IN AC(A31)-2 'MALFUCTION 1 1 IMP I

10043 I i

OF Il-Isj I  !

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!PM:EDURE OF Ch!OFEFATIC R*P BILCW P 8' 15 INSIN:T!0V CT A PRA TICAL i MD-5 IMP 1

'UNSCHEDUdD 10044 I I  !

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!MP (01)-22 10051 1

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i i ATTACHMENT 10 FEBRUARY 9, 1987 SQN DCRDR TRAINING RELATED HECs i

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E__._____________..________.. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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. Ttf A 64 (05 9 69)(OP WP 9 85) y g

{ 7g UNITED STATES GOVERNMENT

• 4

• b O 00 Memorandum TENNESSEE VALLEY AUTHORITY 70: H. L. Abercrombie, Site Director, ONP, O&PS-4, Sequoyah Nuclear Plant FROM: R. Joe Johnson, Director, Division of Nuclear Training POTC DATE: February 9, 1987 -

I

SUBJECT:

SEQUOYAH NUCLEAR PLANT (SQN) - DETAILED CONTROL ROOM DESIGN REVIEW (DCRDR) TRAINING RELATED HUMAN ENGINEERING CONCERN (HECs) i

Reference:

Your memorandum dated November. 17, 1986' (S03 861110 801) ,

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AttachedaretrainingrelatedHECswithEheDivisionofNuc3 ear Training Comments. This should resolve these items. .

1 If more information is desired contact Clyde O. Brewer at 751-0303. l d's u R. Joe Johnson N CHN: COB:ERL Attachments '

cc (Attachments):

RIMS, MR 4N 72A-C '

1 This was prepared principally by C. O. Brewer I i

OFFICE OF PLT MGR CNP SEQUOYAH M D vr snt ose; tor.3 omcE FEB10'B7 ffB l ? '87 g L

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RESPONSE TO HUMAN ENGINEERING CONCERN (HEC) NO. H -0005 To alleviate this concern the following actions have been taken:

1. New emergency instructions have been developed and implemented for operators use at SQN.

2. Training on SQN emergency procedures has been increased in the operator training programs to further increase the operators I understanding and working knowledge of when and hcw to use these procedures. . .

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O RESPONSB TO HUMAN ENGINEERING CONCERN (HEC) NO. XK-0-024 To alleviate this concern the following reasons / actions have been given/tsken:

E

1. In the past plant staff'ing requirements required one or two large classeo. Future plant staffing requirements and personnel available indicates future class size will range from ~ 4-10 trainees per class.

2. Operator Certification Trainin6 Program has been increased from 13 to 16 weeks.

3. Operator Prelicense Training Pro 6 ram has been increased from 7 to 9 weeks.

4. Licensed Operator Requalification Training Program has been increased from 4 to 6 weets annually.

5.

Extension of Licensed Operator Requalification Training Program allows more time to cover annual training commitments and plant systems.

6. Instructors are a'ssigned to a class for its duration unless unexpected circumstances arise.

7.

The majority of the operator training provided is structured to teach the trainees how to satisfactorily operate the plant.

Unfortunately, we do not have control over examinations administered by NRC to trainees so, a certain amount of training has to be devoted to those particular examinations. This provides the trainees with the necessary exposure needed for taking these required examinations.

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RESPONSE TO HUMAN ENGINEERING CONCERN (HEC) NO. XX-0-025 To alleviate this concern the following actions have been taken:

1. A constant effort is being made to concentrate on the essential and actual information required to be covered in Licensed Operator Requalification Training.

2. Licensed Operator Requalification Training Program has been increased from fcur (4) to six (6) weeks annually to more thoroughly cover annual training commitments and plant systems.

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l RESPONSE TO HlTMAN ENGINEERING CONCERN (HEC) NO. IX-0-028 To alleviate this concern the following action has been taken*

.]

The theoretical aspect of training has been restructured to increase training in the area of practical application. l I

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RESPONSE TO HlfMAN ENGINEERING CONCERN (HEC) NO. XX-0-036 Since TVA furnishes NRC with the reference material for use in preparing licensing exams, it is unfortunate that TVA'has no control over how or whethe'r NRC uses this material or who NEC.nssigns to administer the exams. Therefore, this is a valid concern that cannot be alleviated by TVA alone.

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RESPONSE TO HlTMAN ENGINEERING CONCERN (HEC) NO. XX-0-037 1

To alleviate this concern all AOIs are covered in a two year cycle by one of  ;

the following methods: .l

1. Simulator Training

2. Classroom Lecture I

3. Walk-Through  !

4. Sine-off Training Letters '

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. .i RESPONSE TO HUMAN ENGINEERING CONCERN (HEC) NO. IX-0-064 SQN operator training has.always stressed the importance of following procedures to ensure that operational parameters are not exceeded during any manipulation of the plant facilities. Therefore, this is not a valid concern )

i for SQN.

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l RESPONSE TO HUMAN ENGINEERING CONCERN (HEC) NO. XI-0-0094 To alleviate this concern the following reasons / actions have been given/taken:

1. Training for the AOIs.; mentioned (except AOI-35) is not provided on a ,

regular basia (all other AOIs are covered on a two year cycle).

after review of these by SQN Operator Training Group Instructors, it is their opinion that operators should not have any. difficulty in following these procedures. These procedures are written in a manner that is easily followed and provides ample time to solve any potential problems that might arise during their performance.

Therefore, this is not a valid concern for SQN. .,

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2. Training is provided annually on AOI-35 in Licensed Operator l

Requalification Training.

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4 HUMA$i nt,1HEERI!4 C01CERN (HC) WOE EIEC h

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Pla nt Seq uoy ah N.i c le a r Pla nt Unit:

HEC D lb.: xx-0005 1 0 - Simu la te r Da te : 8/ 15/ 8 4 HE* Shori Tit le : Operator'{ecuired to Memori ze an Unreasonable Nutber of

~

Eeer een ev Pro cedure 7tess'. ..

1 oe. . tion:

3 Ch ecklist Item: 01- 32 Rev EEC Identified: Ch ecklis t 3ngrviev ',Qa e st ionna ir e Addit 'ional Analy sis e.

Plant System /Subsy stem: All ,_

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Corp onen t s I.r civ ed ( Uh*IP / Na ce) : Pro cedure s .

.t i Rutan Per for::a rc e Itd a iity Af fec ted (vis ion, hearing, decis ionmaking, etc. ):

i

'Ae co ver v of eeeer v. ,

be tailed De s:rhtion: Some Egerators believe meenrv load for e=errenev

, . pro ce d t: t s is 'too hi trh. Ot her o'oerators disa gree . T ht opinion was e xvre s s e d t hat new pr e ce d tre s s ho ul d hel p .

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Inpat t/Signific an:e e2 tbnc ern (identify how c enc ern relate s to events, nodes, fu n: t ion s, ta sk s, any sa fe ty cor.ceq uen:es, and de s: rib e rela tion ship to any oth er c onc em s a s appropria te) :

[tr:essive res:-j zation vould in crease likelih3od of operator error when ha vin,r t o a ct en t he s e meeerie s . Also, this would increase stress associated v3 'h re ettslifi cation .

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E 34 213.05 1 s-  ;

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• Revision Date IM1AN ENGINEERING CONCERN (EEC) WORKSHEET

'~ Plant: .Sequoyab Nuclear Plant  !

Unit: 1 0 - Simulator Date: I/15784 EEC ID No.: XX -

0 024 (Panel) (Checklist) (Sequence #)

HEC Short

Title:

Training-major changes to training program to imoreve l operator training and reouce training time Location: MCR Checklist Item: 01-39, DQ-A3 How HIC Identified: Checklist In'te rview Questionnaire AdditionalAnalysis l Plant System / Subsystem: N/A

• Components Involved.(UNID/Name): N/A Human Performance Modality Affected (vision, hearing, decision making, etc.):

l l Decision making Detailed

Description:

Need to reduce size of classes, increase lencth of trainine time. Retraining should include more information en the plant systems.

Use the same instructor throughout training program. Train for operating the plant not predominate 1v to pass exam.

Impact / Significance functions, tasks of Concern (identify how concern relates to events, modes, other concerns as, appropriat.e):

any safety consequences, and describe relationship to any EEC WORKSHIET-SnN

Revisien Date 1

hTIAN ENGINEERING CONCERN (EC) WORKSEET

• Plant: Sequoyah Nuclear Plant '

Unit: 1 0 - Simulator Date: E/15784 EC ID No.: XX -

0 025 (Panel) (Checklist) (Sequence f/)

EC Short

Title:

Training-opinions regardine recualification training of Ros and SRos

^

Location: MCR Checklist Item; .

01-40, OQ-A3 How EC Idrntified: Checklist - Interview Questionnaire Additional Analysis Plant System / Subsystem: N/A l

Co:ponents Involved (UNID/Name): N/A i

}

Human Performance Hodality Affected (vision, hearing, decision making, etc.):

- 'l Decision making i s .

Detailed

Description:

There is too much information covered in '

recualification. Either length of trainine should be increased or decrease i

the marnitude of information. Current training rood but trainine should a

include more actual plant events. '

Impact / Significance functions, tasks of Concern (identify how concern relates to events, modes, I other concerns as, appropriate):any safety consequences, and describe relationship to

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Revision Date _ ._

HUMN ENGINEERING CONCERN (EC) VORKSEET

- Plant: Sequoyah Nuclear Plant Unit: 1 0 - Simulator Date: H/15784 EC ID No.: XX 0 -1 iPanel) (Cneckhst) (Sequence #)

EC Short

Title:

_ Classroom training has little practical operational value Location: MCR Checklist Item: 0I-38 -

How EC Identified: Checklist Interviev Questionnai,e Additional Analysis Plant Syste:/ Subsystem: N/A Co::penents Involved (UNID/Name): N/A Human Performance Modality Affetted (vision, hearing, decision making

( Decisien making

,etc.):

Detailed

Description:

The theoretical trainine in many cases is not used in operating the plant, it is required primarily to obtain a license.

Consequently, extensive theoretical trair.ing while needed should be desirn d to be of a more practical nature. e_

Iepact/ Significance functions, tasks of Concern (identify how concern relates to eveots

, modes, other concerns as, appropriate):any safety consequences, and describe relationship to any

I

• i 4

• 1

, Plant: Sequoyah h'uclear Plant ,

I EC ID No. : XX-0.M " /

l {

02$

i

.s Ph(to/Other

Reference:

Attached C e yes no ID No.

'l Suggested Correction (optional).: - I

~

EC Reported By:_

Name (optional) Position (optional)

EC Reported To: _ J. R. Maner .

Team Member Other Person '

Cec.ments : l There were a large rance of opinions regarding the benefit trainine had on job performance. Consensus appeared to be that the theoretical training was not really used extensively on-the-job and it should *~

i be made more practical in order to be more useful.

Consensus was that training was geared mainly toward passing the licensine' enam.

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EC WOESEIT-SQN

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/ Revision Date I

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D JS IN;1NIIF.!N CON;IEN (EIO) L'0?ISEIT

(

,lant: fem--TYAH

. - hlKLE'AR. ft,A14~(

036 Unit: 1 3 - Simulater ,

EC ID No. : U- 0 * "M Date: - ***:"'N (Fanel) (Che:Elist) (Sequence #) lT fE8 F6 EEC Short

Title:

NRO exa-ination neliev fails to test recuired skills and contributes unne:essarily to eterator strere.

Checklist Ite$: 0I-38, 39, 51A - -

~

Le:ation: .

Eev F.I: *dentified: Che:klist Interviev Questionnaire Additienal Analysis Plant Syste:/Subsysta=: _

N/A I Cenpenents involved (UNID/Name): SA l

l l

Buzan Perf er:ance Medality Af f ected (visic,n, hearing, decision raking, etc.):

Operater selfesteen, nenstecific etress, everall jobs qualification De: ailed

Description:

Since everaters state the NRC examination was the

~

priuci;al source of stress in their Ifves, it should be a meanineful exa-inatien, vet se e coersters sav they could cass the exa- vithout kne ine hev to run the clant. - See cements ,

I i

1 i

Inpa:t/ Significance of Cen:ern (identify hev concern relates to events modes, l fun : ens, tasks, any safety consequences, a.nd describe relationship to any ,

ether concerns as appropriate): .a I

(1) Deerater subieeted to unnecessary stress which vou13 be exceeted te j i

affect their perferrance en the job. '

1'

('2) yessibility exists that eteraters are not arorerristelv eualified.

l l

)

(3) If cart of trainine is unneeded except fer NRC exam then time scent {

t I

__for this trairine creates a need fer eere evertice. I

' l (t.) As one coerater said "vou can't chante the trainine protras to make it l

better without chantine the var the NRC eives exa-s."

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. 4 I-1  :

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arE'pupyAN rs ets - st? s2-17 <

Plant: "_ . . ;_ .- Nuclear Plant / Appendix I i HIC ID No.: XX OC < }

036  ;

1 Phetc/Other

Reference:

Attached /l((/ /][/

yes no . LD No.

Suggested Cerrecticn (optienal):

,, t EIC Reperted By: l

• SMIIO ~

Ua:e (opticnal) Posi:1:n (optional)

EIC Reperted To:  ?>4s-Team P.e::er C/ Otner Person E. 7. Sheely C ==ents: 0;erater resrenses indicate thev feel the licensine eye-instien does r not 'T relate well to knowledee recuired for effective en the ieb eerfor ance.

Responses include:

(1) "would aceles and erances be eeed enecch." (VRC exem (a:eles) and runnine elant (orances)). (2) thev just take for etanted we understand flew paths - I don't think they understand flew caths themself".

(3) you are not allowed to answer based unen what you know about the plant and hev it verks -- this is because examiners don't knew eneuch to ask meanineful

• quest 10ns," (4) I haven't seen an exaniner who had seme knowledre of what our plant is like a5d what we have to kncv. All NRC examiners sheuld have seme operational background . . . It is not that the exams are hard, but they are reflections of ignorance," (5) "We don't =ind having to.take an examination on sorething we know is i=pertant, but a lo'; .J vhat they ask makes no sense "

(6) "SRO examiners don't kncv hew the plant works. They will ask you detailed questions about health physics or tomething else because that is their backtround.

Also they will pick single sentences out of context from the pracedure and ask I you about then . . . they v111' lift a sentence from a procedurc, change one verd, and ask you true/ false."

REFEil 70 SQN REY /EW GF WEN //EC- XX-00/7 E-2 C

/

Revision Date t -

HL?.s3 ENOINIIKIN CON ERN (EIC) WCFX$E!IT Plant:

CEQ@YAH

... -:- Nuclear Plant 037 uniti c>-o- o. saul.ter HIC ID No.: XX - 0 - ::: Date: -;: .';; f:: -

(Panel) (Chectitst) (Sequence () l7fE2 F6 KI: Short

Title:

Not eneuth training is provided en the abner =al everating instructions. .

Location: Check. list Ite=: 01-52 He HE: Identified: Checklist (Intervieh Questicrtnaire Additional Analysis

" ' ~

Plant Syste:/Subsysten: -

Corponents Involved (WID/Name):

• s Hu an Perfer:an:e Modality Affected.(vision. hearing, @ecision eak 3 etc..):

Detailed

Description:

If an Ao! it not exceeted to be used, trainine is limited to study of fr. ediate actien for NRC exam and limited training on the simulater fer these sectiens of the precedute that can be sitiulated. Many sections of the AOl's'can not be sir:ulated because the actict.s are cut in the plant. Except fer A02-27 they are not valked through. -

Intact / Significance of Concert (identify how con c.rn relates to events, niedes, fun:tiens , tasks, any safety consequences, and describe relationship to any other concerns as appropriate): .

~~

W Wu-e

/

SEGWYAH Plant: 'n : - Nuclear Plant Ir.C TH No.: .U -W 037 .

f-[

Photo /Other Ref erence: Attsched / / d/

yes no ID No.

Ssggested Correction (optional):

~

HIC Reported By: b* OMITN Nass (optional) Position (optional)

HIC Reported To: VM

'es: Me= der # Other Person

, J. R. Maner (-

Cc=ments: 5 frter te.eer everessed this as a preble .

There were also a number of ce=ents that related to this ' problem.

A01s 21 and 25 vere mentiened as not listing all of the eeufement affseted when veu lese a vital beard.

REFER TO SIN 2Er/EW 0F h'EN HEC- XX-00tg

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Revisien Date t

r.w ts::xttnN: Co:::tr.x -(xt:) woasntti Sc QV0 YAM Plants-

--__- Nm*lcar Plant bh .

tin i t ID- C")-@- Simulator litC ID No.. XX-0  :-- tt ste: . .. :-

(Panel) (Che:klist) (Sequence f) II[f8 Ih HE: Short

Title:

Emphasis stressed in training on the use &'following of procedures '/

Location: All check. list Items 1.EP. 244-F3017 l

How HI: Identifiec: Che:klist Interview - Questicrmatte 6dA1tional Analysh I P l.in t Syste:/ Subsystem: All

• j Corpenents involved (US1D/Name): i hwranFerfer:ancehodalityAffected(vision, hearing.hisionma etc.):

Lier ailed

Description:

The ic11oving of procedures in the performance of

g. various functiens ef the riar.: should be greatly stressed to erevent the

• accide-!z! exceedir.t the ne!- .a1 etera tier.el cere e t ers .

tr:ta:t/Sigt.ificance of Concern (identify how con:ern relates to events, rodes.

fur.: tier.s. tasks. any safety ccesequen:es, and des: ribe relationship to any other cen: erns as apprcpria:e):

e 0

9m

• e e

O O

47 8

9 l

• e t , og .ed

- s SEQucYAH /

Plant: -  :: I-- Nuclear Plant liCC ID No.: ;CC ~~"f'-- .

00(c h Photo /Other

Reference:

Attached /~/ /7/

yes no ID No.

Suggested Correction (optional):

~

EEC Ke; creed By: * ' LIM hate (optional) Positaon (optional)

EI: Reperted To: c* N-~- 'db ,,,

Team P.c=:er / Ocner Persen e Cc= ent s : _ Obtaine:! frc= Cinna Nuclear Plan 's LG nu .ber 244-83017 REFn To sat Renew o;; WEH Nec-xx-00c.17 9

e to

e' *

, Revision Date i

HUMAN ENGINEERING CONCERN (HEC) WORKSHEIT l

PLANT: Sequoyah Nuclear Plant I Unit 1-2-0 Date 7/16/86 EEC ID NO. -

0096 (Panel) (Checklist) (Sequence #)

HEC Short

Title:

Training may be inadecuate for seldom used AOIs.

Location: NA Checklist Item: NA Eov EEC Identified: Additional Analysis Questionnaire 3/3/86 .

Plant Syste:/ Subsystem: Various Co:ponents Involved (UNID/Name): Procedures 1 1

Human Perfor:ance Hodality Affected (vision, bearing, decision making, etc.):

l Ability to use orecedures

• Detailed

Description:

Operator comments on A0!s 8, 9, 31, 32. 33, and 35 indicate that these orecedures may be preceived as beine difficult because of limited train- '

inc. This eeneern mav be acolicable to most seldom used AOIs.

I= pact / Significance of Concern (identify how concern relates to events, c: odes, functions, concerns as tasks, any safety consequences, and describe relationship to any other appropriate):

It is not clear if the imoact is that the opeLa. e tor's ability to use these procedures is coor due to lack of training or_if it is that his cerception of his ability to use the orecedure is low and that the crocedures are sicole encueh that he can figure them out as he goes.

In either case delav and confusion durine-management of uoset conditions [euld be J* oetted.

This could result in the cendit-ion beceming unnecessarily .

severe and contribute to coerator error. __

DNt1 - 0122Q NES 07/19/86

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

k ATTACHMENT 11 WATTS BAR/SEQUOYAH HEC COMPARISON FORM

l Vill be handled under MX-3-111 i

! . e Revision Date i

s-( HUMAN ENGINEERING CONCERN (HEC) WOFXSHEET Plant: Wat ts Bar Nucicar Plant "

Unit: 1 0 - Simulator

' HEC ID No.: M5 3 001 Date: T2/29/84

~ ~

(Fanel) (Checklist) (Sequence f) --

HEC Short

Title:

Stean generator level deviation alarm is not near related controls '

Location: M-5 Checklist Item: 01-11 How HEC Identified: Checklist Int e rview Questionnaire Additional Analysis Plant System / Subsystem:

Corponents Involved (UNID/Name): #

)

Numan Perf ormance Hodality Af fected (vision, hearing, decision making, etc.):

f Detailed

Description:

See short title Iepact/ Significance of Concern (identify how cencern relates to events, modes, functions, tasks, any safety consequences, and describe relationship to any other concerns as appropriate):

O 1

/

see **

P2g ry Watts Bar Nuciser Plant N C iD No.:

8 001 I

)

\(k . '

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Photo /Other

Reference:

Attached / / /~/

yes no ID No.

• i I

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Suggested Correction (optional): )

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HEC Reported By: -

\

I Name (optionsl) Position (optional)

MEC Reported To: E. J. Sheehy Team Member other Person vi Co=ent s : Reperted 1 tire in response to 07-11 -

(

1 -

I l

Sequoyoh Review of att: Dor CRDR HECs Anplitobfe to SQN: Yet . _ No .

7'<<srM 4

j

' Action Token / Reason:

59N HE C - M5-3174 WRITTEN IDfEB EG -

Fe s -

n. I I:

Signature of Review M A - W ^ Dore // 6

//'

ll

.Will be handled under MX-3-111

, Revision Date 1

s-l

\s HUMAN ENGINEERING CONCERN (HEC) WORESELET i Plant: Watts Bar Nuclear Plant i Unit _1 0 - Simulator i

' HEC ID No.: MS 3 001 Date: 12/29/84 (Panel) (Checklist) (Sequence f) --

REC Short

Title:

Steam generator level deviation alarm is not near

~

related contro'Is -

Location: M-5 Checklist Item: 07-11 How HEC Identified: Checklist Interview Questionnaire Additional Analysis 3 Plant System / Subsystem:

Components Involved (UNID/Name): 1

)

l J l P \

Hu=an Performance Hodality Affected (vision, hearing, decisien making, etc.):

{

f Detailed

Description:

See short title i

\ =~.

/

Impact / Significance of Concern (identify how concern relates to events modes, functions, tasks, any safety consequences, and describe relationship to any other concerns as appropriate):

O em

• t' .-

t: Watts Bar Nuclear Plant P1gEqzCtoNo:

8 001 '

t l( .;,

(

Photo /Other

Reference:

Attached ~

/[/ /[/

yes no ID No. *

)

Suggested Correction (optional):

l

(

HEC Reported By:

Name (optional) Position (optional)

H.EC Reported To: E. J. Sheehv 4!

Team Member Other Person Cc=ents:___ Reported 1 time in response to 0T-11 l

~

l Sequoyoh Review of atts Bar CRDR HECs Anoli:cble to SQN: Yet No

)p/cs ttal

)

Action Token /Reoson: - -

59N HEC - M5-3174 WRITTElv ICFEBf ,

Pe 5 .

I

b. Y signo ure of nev\e-er A'EkONW ' __ oor >/- W - W n-

.Will be handled under MX-3-111

, ,, Revision Date s.

( HUMAN ENGINEERING CONCERN (HEC) WORKSEEET

~

Plant: Wntts Bar huelcar Plnnt

• Unit: 1 0 - Simulator ~

' HEC ID No.: M5 3 001 Date: I'2/29/84 '

(Panel) (Checklist) (Sequence d) -

HEC Short

Title:

Steam generator level deviation alarm is not near i

related controls Location: M-5 Checklist Item: 0T-11 How HEC Identified: Checklist Interview Questionnaire Additional Analysis

)

, Plant System / Subsystem:

Cor.ponents Involved (UNID/Name):

J Hur.an Perf er=ance Modality Af f ected (vision, hearing, decision making, etc.):

r Detailed

Description:

See short title

.\ ~.

I l g

1epact/ Significance of Concern (identify hew concern relates to events, modes, functions, tasks, any safety consequences, and describe relationship to any other concerns as appropriate):

e e =

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l ~ %:

i e

r P1gt: Watts Bar Nuclear Plant II:pC'ID No.:

8 001 i

I(k .

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\ .

Photo /Other

Reference:

Attached / / /~/

• yes no ID No.

• Suggested Correction (optional):

I l

I l \

l l

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F.EC Reportec By:

Name (optional) Position (optional)

FJC Reported To: E. J. Sheehy le i

Team Member Other Person

~

i Co= e nt s :__ Reperted 1 tire in rescense to 07-11 Sequoyoh Review of att Bar CRDR HECs

. . . . . - - - . 7'e p s1%l Ar>oli:cble to SQN: Yes . No Action Token / Reason: -

SQht He t - M5-3/ 74 WRITTE/v 19FEB E(, '

Fe s .

O. IF- I Signoture of Review

//

[b A - Oct., //. 7f. frT' U __-_-_ -

.Will be handled under MX-3-111 e

/

Revision ~

Date g.-

HIJMAN ENGINEERING CONCERN (NEC) WORKSHIIT 4

Plant: Wntts Bar Nuclear PInnt Unit: _1 0 - Simulator ~ *

' HEC ID No :. M5 3 001 Date: 12/29/84 (Panel) (Checklist) (Sequence f) -

HEC Short

Title:

Steam generator level deviation alarm is not near l

reinted controls I l

Location: M-5 Checklist Item: 01-11 l How HEC Identified: Checklist Int e rview Questionnaire Additional Analysis l

Plant Syste / Subsystem:

Corponents Involved (UNID/Name):

s h)

Hu.an Perfor=ance Hodality Affected (vision, hearing._ decision making, etc.):

7' l l . Detailed

Description:

See short title l l

=~_

t 6

Impact / Significance of Concern (identify how concern relates to events, modes, functions, tasks, any safety consequences, and describe relationship to any other concerns as appropriate):

i m

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l 1

f ATTACHMENT 12 MARCH 13, 1987 RESTART REQUIREMENT CRITERIA i

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I

( .-f,64 gos ,.s si lop wa s.asi A02 870313 001 P ".., .

UNITED STATES COVERNMENT

'e .

Memorandum TENNESSEE *- .

VALLEY AUTHORI'rY 0  : Those listed FROM  : S. A. White, Manager of Nuclear Power, LP 6N 38A-C __

DATE  : March 13, 1987

SUBJECT:

RESTART REQUIREMENT CRITERIA

References:

(1) C. C. Mason's memorandum to Those listed dated November 26, 1986 (A02 861126 013)

(2) R. W. Cantrell's memorandum to Those listed dated December 23, 1986 (L44 861224 108) l This memorandum supersedes references 1 and 2 and any other previous ,

l instructions relative to restart requirement criteria. The criteria in the attached table have evolved from the original Volume 2 criteria.

Though the requirements are in essence the same as the earlier criteria, these include the lessons learned through implementation at Sequoyah and Browns Ferry. Use of these refined criteria on Sequoyah and Browns Ferry activities has shown that these criteria do meet the test of being easily understood and easily applied.

> These criteria are in effect for all plants until we revise Volumes 2 and 3. After the units have been restored for normal operations, the respective site directors will make a recommendation to me regarding a transition to sole reliance on plant technical specification requirements.

I

.h. i t.

l 70: See list on page 2 f

MCUo' AM WJCa;JL 8U'(T STE Mct0HS CRC ER 17 .

! (I) _

,s is s a i

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%(M i d 'i f

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.ICs to

, $4:3 J

$ ,.[ * , .

• RESTART REQUIREMENT CRITERIA The following criteria shall be used in evaluating whether a particular item must be resolved prior to startup. .,

1 . The item identifies s' specific deficiency which has significant '

.N probability of leading to the inoperability of a system required for startup or opeMitWritiy' the appropriate Technical S, specifications.

2. The item identifies a programmatic deficiency which has a high probability of causing or has caused a specific deficiency which

~

meets No. I above. .

NOTE: To assist in the determination of required for restart

. relative to Technical Specifications as in criteria No. I and No. 2 above, an affirmative answer to any of the following questions requirer consideration of the item for restart based on Technical l - Specification requirements.

i a. Does the item directly and ad'versely affect safety-related ,

equipment function, performance, reliability, or response time?

b. Does the item indirectly and adversely affect safety-related equipment power supply, air supply, cooling, lubrication, or ventilation?

c. Does the item adversely affect primary containment integrity? I

d. Does the item adversely affect secondary containment integrity?

e. Does~ the item adversely affect control room habitability?

, , . Does the item adversely affect systems used to process radioactive waste? -

g. Does the item adversely affect fire protection or fire loads!

h. Does the item adversely affect the ability of a system or

• component to meet its safety function during a design basis event by impacting the seismic analysis, single failure criteria, separation criteria, high energy line break assumptions, or equipment qualification?

1. Are the programs such as Radiological Health Security, Radiological Emerg,ency preparedness, or Quality Assurance which are necessary for safe conduct of operation of the plant adversely affected?

j. If not corrected prior to retart, could it lead to an uncontrolled release ~or sproad of radioactive contamination-( beyond the regulated area. -

- s

'\ .

,.,v

(

-2 ,

3. The item identifies a specific" deficiency that results in a failure to comply with NRC regulations and ho variance has been approved by .

NRC. *

4. .

TVA has committed to NRC to complete the item prior to restart.

5. The item identifies a specific deficiency which has a significant '

probability of leading to a personal injury during plant operation.

6. The item identifies a specific condition which has a forced outage risk (probability I ostage length) during the next cycle in excess of '-

the critical path time to correct the condition prior to restart.

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ATTACHMENT 13 i

REVISED TVA COMMITMENT REGARDING HED 0220 1 1

1

)

f I s TENNESSEE VALLEY AUTHORITY [ k CH ATTANOOG A. TENNESSEE 374ot SN 157B Lookout Place

JUL 141987 U.S. Nuclear Regulatory Commission ATTN

Document Control Desk Washington, D.C. 20555 Centlemen:

In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 .

SEQUOYAH NUCLEAR PLANT (SQN) - AUDIT OF DETAILED CONTROL ROOM DESIGN REVIEW (DCRDR) PROGRAM

Reference:

Letter fec2 TVA to NRC dated November 26, 1986, " Detailed Control Room Design Review Summary Report For The Sequoyah Nuclear Plant Units 1 And 2" In the June 25, 1987 exit meeting for the DCRDR audit, the inspectors identified one Human Engineering Deficiency (HED) that required interim' A corrective action before its scheduled completion date.

This deficiency. HED-0220, was identified by SQN's DCRDR team through review of the plant licensee event reports (LERs) and potentially reportable occurrences (PR0s). SQN has exceeded the technical specification limit for containment pressure of +0.3 pounds per square inch several times during the life of the plant. The f act that the containment pressure indic ator is located on panel M-9, which is behind the main control panel, is a contributing factor to these occurrences.

HED-0220 requires the installation of a pressure-indicating recorder in the horseshoe section of the control panel and is scheduled for completion by the end of the cycle 4 refueling outage.

Pending the completion of HED-0220, SQN management has taken or will take the following interim corrective actions before mode 2 entry on unit 2.

1. Sample containment atmosphere via Surveillance Instruction (SI)-410.1 once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during applicable modes, so as to provide a standing release to vent containment. This will allow venting without delay to control containment pressure.

2. Draft and implement an SI (SI-2.1), which will require the following actions during modes 1 through 4:

a. record of containment pressure once every four hours;

b. trend containment pressure on P-250 trend recorder; and

c. vent containment, in accordance with System Operating Instruction-30.3E, when containment differential pressure increases to between 0.2 and 0.25 pounds per square inch.

sp yd 7N6~# 3^3 P2>

An Equal Opportunity Employer

1 U.S. Nuclear Regulatory Commission gfff( {g jgg

3. Draft and issue a training letter regarding the need to c'losely '

monitor containment differential pressure and take appropriate venting action.

The enclosure lists new commitments made by SQN.

If you have questions concerning this issue, please telephone M. R. Harding at (615) 870-6422.

Very truly yours, TENNESSEE VALLEY AUTHORITY ,

pdi. Gridley, Director Nuclear Safety and Licensing e

Enclosure ':

cc (Enclosure):

Mr. G. G. Zech, Assistant Director for Inspection Programs l Office of Special Projects j U.S. Nuclear Regulatory Commission l Region II l 101 Marietta Street, NW, Suite 2900 l Atlanta, Georgia 30323 l Mr. J. A. Zwolincki, Assistant Director l for Projects l Division of TVA Projects i Office of Special Projects U.S. Nuclear Regulatory Commission 4350 East West Highway EWW 322 Bethesda, Maryland 20814 Sequoyah Resident Inspector Sequoyah Nuclear Plant

. 2600 Igou Ferry Road l Soddy Daisy. Tennessee 37379

i; ,

i 2

ENCLOSURE i

LIST OF COMMITMENTS Pending the completion of HED-0220, SQN management has taken or will take the following interim corrective actions before mode 2 entry on unit 2.

1. . f ample containment atmosphere via Surveillance Instruction (SI)-410.1 once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during applicable modes, so es to provide a standing release to vent containment. This will allow venting without delay to control containment pressure.

' 2. Draft and implement an SI (SI-2.1), which wil1 require the following actionJ during modes 1 through 4:

s. ' record of containment pressure once every four hours;

b. trend containment pressure on P-250 trend recorder; and

c. vent containment, in accordance with System Operating Instruction-30.3E, when containment differential pressure increases to between 0.2 and 0.25 lb/in ,2

3. Dreft and issue a training letter regarding the need to closely monitor containment dif ferential pressure and take appropriate venting action.

i

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t ATTACHMENT 14 t.

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i TVA EMPLOYEE CONCERNS EVAf"4 TION 1

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SAFETY EVALUATIONS REPORT FOR EMPLOYEE CONCERNS REPORT NUMBER: .208.1 (B)

TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR POWER PLANT, UNITS-1 AND 2 DOCKET N0s. 50-327 AND 50-328 CONCERN: OE-QMS-3, "The Progam Plan for control room design review is not adequate."

I. .Subiect Category: Engineering 20800 Subcategory: Human Factors; and Engineering Subcategory 22100 Element: Control Room Design Review Program Plan (208.1B)

II. Summary of Issue Employee concern OE-QMS-3 questioned the adequacy of the Control Room Design Review Program Plan (OE-SEP C2-17) to identify and resolve all human engineering concerns that could significantly affect the safe shutdown of TVA's Sequoyah Nuclear Plant.

Ill. Evaluation The TVA program plan for the CRDR consists of eight main tasks. These are:

o Develop the CRDR plan 1

o Perform an operator experience review I l

0 Survey the main controi room (MCR) and the auxiliary control room  ;

I (ACR) o Perform task analysis o Assess for priority l

l 0 Develop recommendations for corrective . action o Prepare an action plan o Prepare the Summary Report for NRC.

The development of the CRDR plan took place during the period from November 1980,. when NRC issued the TMI Action Plan (NUREG-0737), until the end of 1984. During this period several discussions were held with NRC as

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i described above, and the plan was developed. The final revision to the CRDR plan (0E-SEP-82-17) occurred ih August, 1985 (App. A, 5.e), at which time NRC comments and the lessons learned from having initiated the program were i incorporated (App. A, 5.nn).

1 The actual implementation of the program began August 23, 1983, when a f training cc . x on human factors was conducted for the CRDR team members (App. A,t.ii). The review of operator experience was conducted in several 4 steps. Initially the operators completed a basic questionnaire and the i results were used to develop an addendum to the questionnaire. This new questionnaire was then given to the operators, who the CRDR team felt could best provide answers (App. A, 5.nn).

In addition to using questionnaires, the CRDR team interviewed 17 operators after having first received instruction in effective interview techniques from a consultant human factors specialist.

For the survey of industry experience, the team reviewed an INP0 sort f

! of Licensee Event Reports (LERs), Significant Event Reportss (SERs), and l INP0 Operations and Maintenance Reminders (0MRs) involved either directly or ,

A detailed indirectly with control room design control room operators.

review was also performed of all Sequoyah LERs and reactor trips to identify any with control room or operator involvement. Finally, the results of the CRDR effort at Watts Bar were aiso reviewed for applicability to Sequoyah.

These efforts were completed in March 1986.

Surveys of the MCR and the ACR were initiated in September 1984. The 3 associated tasks of performing a sound survey, a lighting survey, and a i' survey of the HVAC for the MCR were completed during 1984 and 1985 (App. A, 5.x). The MCR/ACR surveys were completed in March 1986.

The CRDR team performeo a task analysis for all emergency operating procedures (EOPs) identified for SQN. The E0Ps, which had been developed by the site E0P team, were analyzed by the CRDR team. The task analysis was completed in April 1986 (App. A, 5.nn).

In March 1986 the Essex Corporation (Essex) reviewed the Sequoyah CRDR documentation (App. A, 5.vv). Essex reviewed only the documentation related to the operating experience review, the control room surveys, and the task analysis, as the remaining portions of the CRDR were not complete. Based on their review, Essex concluded that the SQN CRDR documentation is responsive to the guidelines of NUREG-0700; adequately describes the HECs; and provides a track to data collection methods and NUREG-0700 guidelines. The report summarized the documentation as "... adequate and, when complete, should provide an adequate basis for control room design improvements and for NRC audit." The report also mentioned that Essex would work with the CRDR team to develop additional task analysis information. In addition to reviewing CRDR documentation, Essex has also provided consulting services in essentially all phases of the CRDR since February 1986, including

significant participation in the preparation of the Summary Report (App. A, 7.b). In November 1986, on completion of the Summary Report, Essex provided a summary evaluation of the TVA CRDR program in which they commented.

CONCLUSIONS TVA submited the Detailed Control Room Design Review (DCRDR) Summary Report for Sequoyah Nuclear Power Plant Units 1 and 2, to NRC on November 26, 1986. A preliminary evaluation of the Summary Report was conducted by '

SAIC which resulted in the identification of a number of concerns. In arder to resolve the concerns and evaluate the Sequoyah DCRDR, a pre-implementation audit was conducted from June 22 to June 25, 1987. During ,

the audit, the NRC staff, accompanied by SAIC and Comex representatives, performed a detailed evaluation of TVA's DCRDR. The evaluation included examination of TVA's DCRDR documentation, discussions with the DCRDR study team, inspection of the existing control room, and inspection of mockups and proposed corrective action modifications. This report reflects the consolidated findings and conclusions of the NRC audit team. The conclusionsareprovidedbelow,organizedbythenineSupplementItoNUREf- 4 0737 DCRDR requirements. 1

1. The establishment of the multidisciplinary review team used for the' DCRDR meets the requirement of Supplement I to NUREG-0737.

2. The system # unction and task analysis, which was based on Revision 1 of  !

the Westinghouse Emergency Response Guidelines and supplements, meets the requirements of Supplement I to NUREG-0737.

3. The control room inventory meets the requirements of Supplement I to NUREG-0737,

4. The control room survey methodology and results meets the requirement of Supplement 1 to NUREG-0737.  !

5. The methodology for and results of assessment of human engineering discrepancies meet the requirements of Supplement I to NUREG-0737.

6. It is the audit team's judgment that TVA has conducted an appropriate program for selection of design improvements. It is also the audit team's judgment that none of the DCRDR Category 1 or 2 human l engineering discrepancies are significant enough to warrant delay in l restart or limits on plant operation. However, ir order for TVA to l meet the Supplement I to NUREG 0737 requirement for selection of design improvements, it will be necessary for TVA to submit the confirmatory document described in the DCRDR Technical Evaluation Report.

7. The methodology for verifying that the control room modifications correct the HEDs meets the requirements of Supplement I to NUREG-0737.

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8. The methodology for verifying that the control room modifications do not introduce new HEDs meets the requirements of Supplement 1 to NUREG-0737.

9. The coordination of the DCRDR with other programs, including upgraded E0Ps, SPDS, Reg. Guide 1.97, and training, meets the requirements of Supplement I to NUREG-0737.

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SAFETY EVALUATIONS REPORT FOR EMPLOYEE CONCERNS REPORT NUMBER: 208.1 (B)

TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 AND 2 i DOCKET N0s. 50-327 AND 50-328 i CONCERN: WI-85-100-007, XX-85-122-020, XX-85-122-021, XX-85-122-022, "Sequoyah: Human Factors Engineering and/or reviews have not been implemented for control panels and stations."

I. Sub.iect j l

Category: Engineering 20800 l Subcategory: Human Factors; and Engineering Subcategory 22100 ,

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Element: Control Room Design Review Program Plan (208.1B)

II. Summary of Issue l

Human factors engineering and/or reviews have not been implemented for I control panels and stations. CI expressed that this is a violation of j NUREG-0700. CI further stated that there are too many poor engineering l practices in this area. CI has no further information.

CONCLUSIONS ,

TVA submited the Detailed Control Room Design Review (DCRDR) Summary l Report for Sequoyah Nuclear Power Plant Units 1 and 2, to NRC on November  !

26, 1986. A preliminary evaluation of the Summary Report was conducted by l SAIC which resulted in the identification of a number of concerns. In order l to resolve the concerns and evaluate the Sequoyah DCRDR, a pre-l implementation audit was conducted from June 22 to June 25, 1987. During the audit, the NRC staff, accompanied by SAIC and Comex representatives, performed a detailed evaluation of TVA's DCRDR. The evaluation included j examination of TVA's DCRDR documentation, discussions with the DCRDR study team, inspection of the existing control room, and inspection of mockups and l proposed corrective action modifications. This report reflects the l consolidated findings and conclusions of the NRC audit team. The conclusions are provided below, organized by the nine Supplement 1 to NUREG-0737 DCRDR requirements.

l 1. The establishment of the multidisciplinary review team used for the DCRDR meets the requirement of Supplement I to NUREG-0737.

2. The system function and task analysis, which was based on Revision 1 of the Westinghouse Emergency Response Guidelines and supplements, meets the requirements of Supplement I to NUREG-0737.

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l 3. The control room inventory meets the requirements of Supplement I to NUREG-0737.

4. The control room survey methodology'and results meets the requirement of Supplement 1 to NUREG-0737.

5. The methodology for and results of assessment of human engineering discrepancies meet the requirements of Supplement I to NUREG-0737.

6. It is the audit team's judgment that TVA has conducted an appropriate program for selection of design improvements. It is also the audit team':, judgment that none of the DCRDR Category 1 or. ? human engineering discrepancies are significant enough to warrant delay. in restar' or limits on plant operation. However,' in order for TVA to meet tl e Supplement I to NUREG-0737 requirement for selection of design improvements, it will be necessary for TVA' to submit the two (

confirmatory document described in the DCRDR Technical Evaluation Report.

7. The methodology for verifying that the control room modifications I correct the HEDs meets the requirements of Supplement 1 to NUREG-0737.

8. The methodology for verifying that the control room modifications do l not introduce new HEDs meets the requirements of Supplement I to NUREG-0737.

9. The coordination of the DCRDR with other programs, including upgraded i E0Ps, SPDS, Reg. Guide 1.97, and training, meets the requirements of Supplement 1 to NUREG-0737.

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ATTENDANCE ROSTE' R

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October 10, 1986 l Date: .

1 Meeting: CRDR/EOP Chairman : D- J. Martin ,

Recorder: J. Martin Address Organization Division Name (if TVA) (or Firm) and ?> ranch .

3. Martin V8 D207 C-K TVA CRDR Team Leader -

G. Terestra SON-operations TVA EOP Team Members

c. Sr-irk 1ned " " ei n n ,,

J. R. Walker E0P & CRDR Team rianager l

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