ML19325C147

Post-Implementation Audit Rept for CP&L Brunswick Steam Electric Plant Units 1 & 2 SPDS 890515-18.

ML19325C147
Person / Time
Site:	Brunswick
Issue date:	07/07/1989
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML19325C146	List: ML19325C145 ML19325C147
References
CON-FIN-D-1311 SAIC-89-1131, TAC-M51226, TAC-M51227, NUDOCS 8907190139
Download: ML19325C147 (62)

Text

,

, ENCLOSURE 2 l

SAIC 89/1131 POST-lMPLEMENTATION AUDIT REPORT

• FOR CAROLINA POWER AND LIGHT COMPANY'S BRUNSWICK STEAM ELECTRIC PLANT UNITS 1 AND 2 SAFETY PARAMETER DISPLAY SYSTEM MY 1518,1989 TAC N05. M51226, M51227 July 7, 1989 -

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l Prepared for:

U.S. Nuclear Regulatory Comission Washington, D.C. 20!$5 Contract NRC 03 87 029 Task Order No. 36 f O

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TABLE OF CONTENTS I 1

Section Egg l.0 INTRODUCTION............................................... I BACKGR0VND.................................................

1 2.0 I  !

2.1 Summary of Requirements............................... 1  !

l 2.2 Milestones in Completion of Brutiswick SPOS l Requirements.......................................... 3 l 2.3 System Description.................................... 4 )

3.0 EVALUAT10N................................................. 4 I

, 3.1 Concise Display of Critical Plant Variables to Control t Room Operators........................................ 4 ,

3.2 Located Convenient to Control Room Operators. ... . . .. .. 5  ;

3.3 Continuous Display of Plart Safety Status  ;

Information........................................... 5 ,

3.4 Should Have a High Degree of Reli1.bility. . . . . . . . . . . . . . 6 l 3.5 Suitably isolated From Electrical and Electronic t Interference with Safety Systems...................... 8 3.6 Designed Incorporating Accepted Human Engineering -

Princip1Ls............................................ 8 3.7 Minimum Information Displayed Should Be Sufficient To Determine Safety Status With Respect to Five  :

Functions............................................. 10 3.8 Procedures and Operator Training Addressin Actions With and Without SPDS.....................g............ 10 l r

. 4.0 CONCLUS10NS................................................ 11

5.0 REFERENCES

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I Attachment 1 Audit Agenda l

l Attachment 2 List of Audit Participants l t l Attachment 3 Licensee Presentation on ERFIS and SPDS Attachment 4 Licensee Presentation on SPDS Operating Experience i i

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l l POST lMPLEMENTATION AUDIT REPORT FOR j CAROLINA POWER AND LIGHT COMPANY'S BRUNSWICK STEAM ELECTRIC PLANT UNITS 1 AND 2 SAFETY PARAMETER DISPLAY SYSTEM '

MAY 15 18, 1989 ,

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1.0 INTRODUCTION

This report documents the findings of a post implementation audit of j the Safety Parameter Display System (SPDS) at Carolina Power and Light Company's Brunswick Steam Electric Plant. The audit was conducted by the Nuclear Regulatory Commission (NRC) during a site visit May 15 18, 1989. It i was conducted to determine the status of the Brunswick $PDS with regard t o, the eight minimum requirements of NUREG 0737, Supplement 1 (Reference 1).

The audit team consisted of an NRC team leader and contracters from )

Science Applications International Corporation (SAIC) and Comex Corporation, j The team consisted of specialists in human factors engineering and nuclear., l power plant operations. i The audit agenda is provided in Attachment 1. A list of audit meeting attendees is provided in Attachment 2. The licensee's SPDS presentation f

packages are provided in Attachments 3 and 4.  !

2.0 BACKGROUND

2.1 Sumary of Recuirements The principal purpose and function of the SPDS is to aid control room personnel in determining the safety status of the plant and in assessing t whether abnormal conditions warrant corrective action by operators to avoid j degradation of the core. The SPDS can be particularly important during anticipated transients and the initial phase of an accident.

All holders of operating licenses :nust provide an SPDS in the control  !

l rooms of their plants. The NRC requirements for the SPDS are defined in NUREG 0737, Supplement 1. This document requires licensees and license l

l applicants to prepare a written safety analysis report sufficient to assess '

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' the safety status of each identified function for a wide range of events, including symptoms of severe accidents. Licansees and applicants must also '

prepare an implementation plan for the SPDS that contains schedules for design, development, installation, and full operation of the SPDS as well as a design verification and validation p10 The safety analysis report and the implementation plan are submitted to the NRC for staff review. The results of the staff's review are published in a safety evaluation report.

The SPDS requirements, as defined by NUREG 0737, Supplement 1, are:

1. Should provide a concise cisplay of critical plant variables to control room operators (Paragraph 4.1.a).

i. Should be located convenient to control re0m operators (Paragraph 0

, 4.1.b). )

3. Should continuously display plant safety status information (Paragraph 4.1.b).

4. Should have a high degree of reliability (Paragraph 4.1.b). [

5. Shall be suitably isolated from electrical or electronic )

l interference with safety systems (Paragraph 4.1.b).

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6. Shall be designed incorporating accepted human factors engineering i principles (Paragraph 4.1.e).

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t 7. Minimum information displayed shall be sufficient to determine '

! plant safety status with respect to five safety functions

(* (Paragraph 4.1.f): '

i. Reactivity control

11. Reactor core cooling and heat removal from the primary system iii. Reactor coolant system integrity tv. Radioactivity control

v. Containment conditions.

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8. Procedures and operator training addressing actions with and l without SPDS should be implemented (Paragraph 4.1.c). l Guidance for evaluating the implementation of the above requirements is i provided by Appendix A to Section 18.2 of NUREG 0800 (Reference ?) and other ]

l documents cited therein, particularly NUREG 0700 (Reference 3).

In 1985, an NRC survey of six operating Safety Parameter Display Systems was performed to sample the status and quality of the systems. The survey included on site evaluations of licensee documentation.and hardware, as well as interviews with operations personnel. The survey findings, including descriptions of major deficiencies, were identified in Inspection J and Enforcement Information Notice No. 86 10, " Safety Parameter Display. l System Malfunctions ' dated February 13, 1986 (Referance 4).

On April 12, 1989, the NRC issued Generic Letter 89 06: ' Task Plan Item i 1.D.2 Safety Parameter Display System 10 CFR 50.54(f)* (Reference 5).

The generic letter requests all licensees of operating plants to perform an ,

assessment and evaluation of the implementation status of their Safety Parameter Display System. In addition, NUREG 1342, "A Status Report Regarding Industry Implementation of Safety Parameter Display Systems," '

dated April 1989 (Reference 6), was included with the generic letter to assist licensees in implementing SPDS requirements.

2.2 Milestones in comoletier 'f_ Brunswick SPDS reouirements Carolina Power and Light Company submitted a Safety Analysis Report for  ;

the Brunswick Steam Electric Plant SPDS by letter dated December 27, 1984 i (Reference 7). The licensee responded to an NRC request for additional information about isolation devices in a letter dated April 19, 1985 (Reference 8). The NRC issued a safety Evaluation Report in May 1985 (Reference 9). A request for additional information was enclosed with that report. The licensee responded to the NRC's request for additional ,

information by letter dated July 19, 1985 (Reference 10). The NRC issued a final Safety Evaluation Report for the Brunswick SPDS in November 1985 (Reference ll).

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The Brunswick SPDS was declared operational by the licensee in December 1988. A post implementation audit was conducted during the week of May 15, 1989, as documented in this report.

2.3 System Descrintion The Brunswick SPDS is a subsystem of the Emergency Response Facility Information Systen.(ERFIS). The ERFIS is implemented on redundant, Digital ,

Equipment Corporation VAX 11/785 microcomputers. Signals are provided to  !

the system from three sources: the plant process computer (Honeywell 4010),

the Suppression Pool Temperature Monitoring System (SPTMOS), and the i Meteorological Tower computer. The system is a modification to the General j

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Electric GESSAR 11 system which has received NRC review and approval. The.  !

SPDS terminals are Toshiba Intelligent Graphics Display Terminals.

l The Brunswick SPDS displays were developed using the baseline General Electric GESSAR 11 system displays. They were updated and customized to more effectively support the plant specific emergency operating procedures. ,

3.0 EVALVATION ,

W The NRC audit team evaluated the SPDS against the eight NUREG 0737,  ;

Supplement I requirements. The audit findings follow.

3.1 Concise Disolav of Critical Plant Variables to Control Room Ooerators The evaluation of the concise display criterion included a review of physical location of displayed information and technical information '

organization within the display screens. Both physical display grouping and technical inform 6 tion organization are necessary to judge whether the i

display is concise.

All entry conditions of the Brunswick emergency operating procedures (EOPs) and the critical safety function status information required by NUREG 0737, Supplement 1, are displayed in a strip of 20 status boxes called the ' plant status matrix " in Ndition, several key, E0P decision making l variables (e.g., ' Safety Relief Valves Shut") are included in the matrix.

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. '}. , .l p The plant status matrix appears at the bottom of all SPDS display screens (approximately 75 total screens available).

It was the review team's judgment that the licensee met the NUREG 0737, Supplement I requirement for a concise display of critical plant variables.

3.2 Located Convenient to Control Room Ooerators The licensee has ident i' w the primary control room users of the SPDS to be the Shift Foreman and ' N .. ft Technical Adv;sor. The Shift Foreman is the control room nianager ;f emergency response and is assisted in this r le by the Shift Technical Advisor. The Brunswick Unit I and Unit 2

& control rooms each have two dediccted displays, one at the Shift Foreman's desk in front of the center of the control boards, and a second on the vertical portion of the control boards just to the right of the primary

- reactor and turbine control board section. The desk-mounted display is on a platform that can be rotated horizontally and tilted up and down. The existing display locations are convenient to the primary users and also to ,

the control board operators. The licensee informed the NRC review team that they plan to add two more terminals to each control room. The present configuration is satisfactory; the additional terminals should make the SPDS

__ even more convenient to access.

lt wei the review team's judgment that the licensee met the NUREG-07'47, I requirement for a display located convenient to control

- Supplement room l E- operators.

3.3 Continueus Disolav of Plant Safety Status Information l I

< The required phnt safety status informa: Mn is povided in the plant k status matrix which appears at the bottom o' each SPDS display screen. y

= Thus, no matter which SPDS display i: retrieved, the critical informat'on is j present. There is, however, the potential to defeat the continuous di spl ay i

. of critical plant safety status information, as explained below. I

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Many other, non-SPDS displays can be accessed by shifting the Toshiba l

._ terminals to the VT100 mode. ihe Vi 0 displays do not incluoe the plar status matrix that tppears on all SPDS displuys. At the time of the audi+,

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j control room personnel were not knowledgeable of the keyboard manipulations  !

nectssary to put the control room terminals in the VT100 mode. However, the I licensee plans to train control room personnel to use the VT100 mode, thus personnel may learn how to enter this mode even before they are trained to use it. There is no physical guarantee that at least one terminal in each control room will be maintained in SPDS mode at all times.

It was the review team's judgment that the licensee has not met the NUREG 0737, Supplement I requirement for a continuous display of plant l safety status information because non SPDS displays can be accessed and displayed on the terminal thus eliminating the SPDS.

3.4 Should Have a Hioh Dearea of Riliability '

The NRC defines a reliable SPDS as one in which the computer hardware and software function with better than 9P% reliability. The Central Processing Unit (CPU) availability of the Brunswick SPDS is well documented l to exceed 99%. However, overall system reliability, both as observed by the .

audit team and as described by system users, is much lower than 99%.

The licensee identified to the audit team the various system problems which are degrading the reliability of the Brunswick SPDS in a document titled " Operating Experiences" (see Attachment 4). The problems include:

o Software problems o Database problems o Equipment problems o Additional training needs o Emergency Opertting Procedure utilization o Historical data utilization Specific examples of problems identified by the audit team are summarized below:

o four to five of the 20 plant status matrix boxes were intermittently indicating invalid data.

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I o The frequency of terminal lockups, although improving, ranges from -l once per shift to three times per week (per discussion with system usars).  ;

o The terminal used during user interviews had some unexplained i missing displays.

o Two of the control room terminals were noted to be locked up for about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the first day of the audit.

o Operators and Shift Technicil Advisors interviewed had a relatively low level of confidence in the SPDS as it operates in the control room. They had, however, a strong appreciation of the potential value of the SPDS based on their experiences with it in simulator training. (In the simulator, the SPDS is highly reliable, since it is driven from the simulator computer rather that from hundreds of separate plant data inputs.)

, t Additional concerns identified that relate to system reliability are:

o At 100% steady state power operation of the Brunswick units, and with routine control roem queries of the system in pro cess, the l CPU loading is about 75%. This does not allow sufficient excess capacity for periods of greater than routine demands on the system. Users reported that display access time (response time to keyboard manipulations) approximately doubles during transients or periods' of heavy use of other terminals attached to the system (e.g., terminals in the Technical Support Center and Emergency Operations Facility). Some displays take as long r.s 60 seconds to

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access during these higher use periods, o Problems with the system print-queue manager lead to CPU failures i that can affect the respnse time and the redundancy of the system.

L lt was the review team's judgment that the licensee has not met the +

NUREG 0737, Supplement I requirement for a high degree of reliability. The licensee has defined an action plan to resolve the reliability problems of 7

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,. .' , '*l the SPDS and has started work on a number of the tasks involved (see Attachment 4, final page). The licensee stated that they currently estimate that the corrective actions needed to resolve these problems will be i completed by the end of 1990. J 3.5 Suit ably Isolated from Electrical and Electronic Interference with Safety Systems The NRC issued a safety evaluation report (SER) to the licensee on May 8, 1985, approving the isolation devices that interface with the SPDS. This SER was based on an assumption that the Brunswick installation of the generic General Electric GESSAR II would contain only the isolation devices  ;

described in General Electric Topical Report NEDE 30284P. During the audit,-

it wa. noted that the Suppression Pool Temperature Monitoring System

, (SPTMOS) is not part of the standard General Electric design and that the associated isolation devices may require separate review by the NRC.

4 The signals to the ERFIS system from SPTMOS appear to be the source of <

many of the invalid data alarms on the SPDS. The licensee stated that the" f signal problems are being caused by an electronic switch in the SPTMOS data l

link.

The isolation devices used ir :PTMOS are Viitaker/Exosensor Model -

117B028. These isolation devices are not on the list of approved isolators contained in NUREG 1342. During the course of the audit, the licensee was not able to locate testing data for the isolators to demonstrate that they -

are acceptable with respect to maximum credible faults, electrical interference, and environmental and seismic qualification.

g It was the audit team's judgment that the licensee no longer meets the huREG 0737, Supplement I requirement for isolation of the SPDS from safety l systems because isolation devices that had not been reviewed by the staff l were installed on the suppression pool temperature monitoring system.

1' l 3.6 Desianed Incorooratino Accented Human Enoineerino Principles ly l The NRC staff confirmed in the November 1985 safety evaluation report of the Brunswick SPDS that human factors engineering was an integral part of 8

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the licensee's design process. During the May 1989 audit, the review team was able to evaluate the human factors engineering of the completed SPOS.

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The Brunswick $PDS is based on the NRC-approved General Electric GESSAR -

11 system, which had acceptable human factors design. Modifications made to  ;

the generic design by the licensee were found to have improved the human j factors design of the system. Examples of improvements include: (1) the removal of extraneous titles from displays; (2) the addition of plant- l specific status boxes such as ' Radiation Release'; and (3) customization of displays and supporting algorithms to the plant-specific, current revision i E0Ps.

With regard to sechnical content, the critical safety function information is directly integratea with the licensee's E0Ps. The SPOS

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displays the information needed to assess the safety functions, alert the l

shift-foreman to changes in safety function / E0P entry condition status, and

support E0P performance. As shown in Attachment 3, Figure 18, with regard I to technical information organization, the displays are hierarchically ,

organized according to the structure of the plant-specific E0P set.

Although the human factors engineering of the Brunswick SPDS was found I to be generally acceptable, the following concerns were identified by the ,

audit team:

l o The licensee identified several potential ERFIS system HEDs and had not yet assessed their significance or designed corrections, o The ERFIS has no screen message (e.g., WAIT - PROCESSING) to alert a terminsi operator that the system has received a keyboard instructicn satisfactorily and that it is being processed. The lack of such a message can cause less experienced operators to I

continue entering commands, believing that the system is not processing their instructions. Repetitive keypunching can cause system lockups.

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o The slow response to commands (5 to 6 seconds for simple screens l>

and 28 to 30 seconds for complex screens) during steady state operation detracts from the usefulness of the system. As 1 _

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.. i previously discussed, these response times can be doubled during transients or when all 11 ERFIS terminals are in use.

It was the review team's judgment that the licensee has not met the NUREG 0737, Supplement I requirement for a display system design incorporating accepted human factors engineering principles.

3.7 Minimum Information Disclaved Should Be Sufficient To Determine Safety Status With Resnect To Five Functions The Brunswick SPDS displays provide status information for all of the plant safety functions listed in NUREG 07?7, Supplement 1. The SPDS Plant Status Matrix alarms are driven by algorithms which are a faithful i

. representation of the plant specific E0Ps. All of the parameters l recommended for monitoring safety function status of Boiling Water Reactors .

are included except Source Range Monitor data.

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Source range monitors are necessary to mcnitor reactivity status during ,

shutdown and startup. The combination of average power range monitors and source range monitors provide sufficient overlap to cover the intermediate

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l range. Therefore the average power range monitors and source range monitors l represent the principle SPDS neutron flux indicators for reactivity control.

It. was the audit team's judgment that the licensee did not meet the NUREG-0737, Supplement I requirement for display of minimuin information sufficient to determine safety status with respect to five functions because

.the SPDS did not display source range monitor information.

3.8 Procedures And Ooerator Trainina Addressina Actions With And Without SED.S The licensee has developed and implemented procedures and training for use of the SPDS. The procedures and training material were reviewed during the adit and found to be appropriate. All control room personnel have received simuistor training in the use of the SPDS. The licensee stated that they have not as yet conducted training specifically directed tr euergency operations without SPDS. Their reason for this is that the syr-was only recently introduced, so that the immediate need is to learn how *'

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. 1 was only recar.tly introduced, so that the immediate need is to learn how to l operate with this aid available. At the time of the audit, no formal plan i existed to conduct simulator training with the SPDS unavailable.

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1 It was the audit team's judgment that the licensee has not met the NUREG-0737, Supplement I reouirement to provide training addressing actions '

with and witirout SPDS, because the licensec did not have a formal plan to train for operations without SPDS.

4.0 CONCLUSION

S The NRC audit team conducted a post-implementation audit of Carolina Power and Light Company's Brunswick Stean Electric Plant Safety. Parameter.

Display System on May 15-18, 1989. 1he purpose of the audit was to determine the status of the SPDS with regard to the minimum requirements of NUREG-0737, Supplement 1.

Based on the on site audit, it was the NRC audit team's judgment that ,

the licensee has not met the following NUREG 0737, Supplement 1 SPDS criteria:

o Criterion 3: Continuous Display of Plant Safety Status Information (Section 3.3) >

o Criterion 4: High Degree of Reliability (Section 3.4) o Criterion 5: Suitable Isolation (Section 3.5) o Criterion 6: Human Engineering (Section 3.6) o Criterion 7: Minimum Information To Determine Safety Status (Section 3.7) o Criterion 8: Procedures and Operator Training Addressing Actions With and Without SPDS. (Section 3.8) 33 l

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5.0 REFERENCES

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1. NUREG 0737, Supplement 1, Requirements for Emergency Response 1 Capability, Generic Letter 82 33, USNRC, December 17, 1982.

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2. NUREG 0800, Standard Review Plan of Safety Analysis Report for Nuclear Power Plants, Section 18.2, Rev. D Safety Parameter Display System ,

(SPDS), Appendix A to SRP Section 18.2, USNRC, November 1984.

3. NUREG 0700, Guidelines for Control Room Design Reviews, USNRC, September 1981.

4. IE Information Notice No. 86-10: Safety Parameter Display System' Malfunctions, USNRC, February 13, 1986.

S. Generic Letter 89 06: Task Action Plan Item I.D.2 - Safety Parameter ,

Display System 10 CFR 50.54(f), USNRC, April 12, 1989.

6. NUREG-1342, A Status Report Regarding Industry Implementation of Safety Parameter Systems. USNRC, Arpil 1989.

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7. Letter from E.E. Utley (CP&L) to D.B. Vassallo (NRC) with enclosures dated December 27, 1984.

8. Letter from S.R. Zimmerman (CP&L) to n.B. Vassallo (NRC), dated April p 19, 1985.

l L 9. Letter from D.B. Vassallo (NRC) to E.E. Utley (CF'L), dated May 16,

!. 1985.

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10. Letter from A.B. Cutter (CP&L) to D.B. Vassallo (NRC), dated July 19, 1985.

l l 11. Letter to E.E. Utley (CP&L) from D.B. Vassallo (NRC), dated November 20, 1985.

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{ L ATTACHMENT 1 t f'

AUDIT AGENDA )

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..=.. .- l g ',;' j TENTATIVE AUDIT AGENDA FOR i

SAFETY PARAMETER DISPLAY SYSTEM AT CAROLINA POWER AND LIGHT COMPANY'S BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 MAY 15 THROUGH 18, 1989 Rav 1 _May 15. 1989 1:30 pm NRC Entrance Briefing and Detailed Control Room Design Review (DCRDR) briefing -

2:00 pm Licensee briefing on DCRDR program at Brunswick 2:45 pm Audit team control room walkdown of the Reactor Pressure Vessel Control or Radioactivity Release Control procedure (Access t o' control room needed; 2 licensed operators needed).

4:30 pm- Audit team documentation of sample human engineering discrepancies ,

identified during control room walkdown 5:00 pm End Day 1 P

Day 2 - Tuesday. May 16. 1999 1

8:00 am Presentation to licensee of findings from E0P walkthrough identified by audit team 8:30 am Review of implemented and proposed DCRDR related control room  ;

modifications including:

l' o Annunciator project modifications i o E0P instrumentation project o Indicator upgrade project

! o Component relocation project 11:00 am Review of schedules for implementating any remaining safety significant HEDs.

11:30 am Review of coordination of DCRDR modifications with changes made in other programs including:

o Safety Parameter Dispicy System l o Regulatory Guide 1.97 instrumentation o Upgraded E0Ps o Operator training

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' 12:00'

' Lunch.

1:00 pm' Licensee Yiscussion of how audit team's sample findings identified during E0P walkdawn, were identified in their DCRDR.

,2:00 pm' NRC' Caucus 3:00 pm NRC and license 9 technical issue discussion and resolution 5:00 pm End day 2 Day 3 - Wednesday.-Mev 17. 1989 -t 8:30 am NRC Safety Parameter Display System (SPDS) entrance briefing ,

l o SPDSs Generic Letter Ste.tus o Frevious NRC-findings rcgarding Brunswick SPDS 9:00 am Licensee briefing on SFDS program results to date 10:00 am SPDS Evaluation (Access to control room and TSC needed) .

1. Parameter selection o Reactivity control  ;

o Core cooling and heat removal from the primary system o Reactor coolant system integrity  ;

Radioactivity control

.o o Containment conditions ,

2. Continuous display of top level safety function information I

3. Concise display of safet,v function information

'4 . Located convenient to contrci room operator High degree of reliability 5.

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6. Design incorporating human factors engineering

7. Procedures and training for SPDS operation u

8. F.'ectrical isolation s 9

17:30 Lunch 1:30 pm Operatorinterviews(S/S;STA;TRGinstructor) 3:00 pm NRC Caucus e

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4:00 pe NRC/ Licensee SPDS technical issues discussion and resolution s

5:00 pm End day 3 I

Day 4 - Thursday. May 18.1989  ;

8:00 am NRC/ Licensee discussion and resolution of DCRDR or SPDS issues (as necessary) ,

. 10:30 am NRC/ Licensee management exit t

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ATTACHMENT 2 LIST OF AUDIT ATTENDEES i

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1 ENTRANCE MEETING i Attendees 5/15/89 H&t![ ORGANIZATION Sam Strickland OPS  :

James Bongarra NRC/NRR/DCPQ Gary Bethke NRC - Comex <

Barbara Par more NRC - SAIC (human factors) <

William H. Ruland NRC - SRI C.f. Blackmond, Jr. CP&L MGR - OPS William B. Geise Project Special, - Simulator Support '

George Barnes CP&L Operations Mike Williams BTU

• Mike Sawtschenko CP&L Cperations

• Mike Beck OPS Ralph Sanders . CP&L NED, Raleigh Randy Weiss CP&L NED, Raleigh Michael J. Pastva, Jr. CP&L Regulatory Compliance ,

Arnold W. Schmich CP&L NFS,' Raleigh Wilbert May CP&L NED, Raleigh .

l T.H. Wyllie CP&L - BNP 1

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1 DCRDR/SPDS EXIT Attendees 5/18/89 r

HAbi ORGANIZATION G.W. Bethke NRC - Comex Barbara Paramore NRC - SAIC Mike Beck OPS , ,

Mike Williams TRANG Mike Sawtschenko OPS George Barnes OPS  ;

Gene Eagle CP&L - TS Ralph Sanders CP&L, Raleigh K.E. Enzor

• W. Levis NRC .

Joe Holder. CP&L T.H. Wyllie CP&L W.B. Geiss CP&L  ;

J. O'Sullivan CP&L  ;

Walt Simpson CP&L ,

Steve Callis CP&L David Dorsett CP&L David Rudoff QA/CP&L Arnold Schmich CP&L/NFS Michael Par,tva CP&L Albort May CP&L C.F. Brackmon, Jr.

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R.E. Helme CP&L ,

James Bongarra NRC/NRR g , , , , , - - - , , , , - ~ , -,,-....,--e ,.n,, ---,. - , - . , _ , - - - ,,,, . - - , ~ . . -- a

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I ATTACHMENT 3 LICENSEE PRESENTATION ON ERFIS AND SPDS i

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ERFIS  ;

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OVERVIEW .

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0:30 - 9:00 ERFIS and SPDS OVERVIEW By Gene Eagle l Computer Support o ERFIS Overview

• o SPDS Overview i o SFDS Generic Letter Status ,

o Previous NRC Findings Regarding

' Brunswick-SPDS '

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l 9:00 - 10:00 SPDS Program Results to Date By- Randy Weius '

Nuclear Engineering Department (NED) l l

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H EMERGENCY RESPONSE FACILITY INFORMATION SYSTEM (ERFIS) '

. . I PURPOSE l

g Monitor plant systems, make plant status and transient '

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f' information.available to operations and other groups, provida ,

l L an SPDS that supports the 20P's, and provide guidar;ce l'

L information'during energency events. ,

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ll SAFETY PARAMETER DISPIAY SYSTEM (SPDS)  ;

l A subsystem of ERTIS.

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, RIN ATTRIBUTES OF BRUNSWICK ERTIS

- Twin Systems (Unit 1 and Unit 2)

- DEC VAX Computer Based

- Toshiba Color Graphics (1024 X 768 Pixels)

- Analogic DAS

-- GE ERTIS Software .

- Computers (2 VAX 11/785, 1 MicroVAX II) per Unit E -

Fiber optics o

- High Hardware Point Scan Rates Analog 250 Times /Sec Digital 500 Times /Sec 1-

- scan, Log, Alarn, Display, Trend Trarisient Analysis M-M-

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ERF15 FUNCTIONS E ERFIS performs the follow' a functions:

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b. System W >'le Frocessing

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d. Technical Support Center (TSC) and Emergency Operations Facility -  !

(EDF) Support

e. .CRT Trending and Plotting

f. Monitoring and Alarming '

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g. Logging-l h. Console Functions 1

1. Editor Console Functions

i. Historical Data Stefage and Retrieval

k. Alare Display ',

1. Graphic Displays -

m. Configuratinn Control s

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VAX/VMS Operating Systee Systen DEC MABITAT Data Base Management System System ESCA CDCI Common Data & Control Interface Application GE Transient Recording & Analysis Application TRA GE RTAD Real Time Analysis & Display Application GE PLA Process Log and Alare Application GE ,

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• STAR'NP, CONSOLE [TS C1 ) 0,0 23-MAY-1985 10:53:55 4.....................................................................#

MAIN MENU

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(_) NELP * .

4.....................................................................

(_) SYSTEM STATUS

(_) APPLICATIDN EXECUTION

[_) APPLICATION DATA BASE EDIT l ) DATA / ARCHIVE /RILOAD -

!_) MAIN DATA BASE EDIT

(_) SYSTEM CONTROL TAB TO THE DESIRED TUNCTION, AND PRESS KEY PAD 2 " SELECT" EXIT SY TYPING */ EXIT \" ON THE COMMAND LINE AND PRESS l ' RETURN" '-

! (this is the e.est. age line) l (this is the HABITAT message line) l ERTIS MAIN MENU

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Figure B Compotar Suppert Group proposed Full Support Organisetten -]

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. Reports to Restter tagineering Supervisor is.the Nuclear group, Technical Support PROJECT ENGINEER

• CONPVIER l

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SENIOR SPECIALIST . SENIOR SPECIAL!ff = i COMPl/71R MARN ARE COMPVITR SOT!VARE '

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i COMPUTER g C PER A Tc A I nn ,

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! Computer Computer Computer -

Tech Tech Ta'ch Systems Supported Unit 3 Process Computer Unit 1 ERTIS Unit 2 Process Computer Unit 2 ERF15

• TBC/E0T ERTis Process Computer Replacement l

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Development of SPDS Displays

- Working Group formed to Modify GE SPDS Displays p m m op = = e

- Overall Modifications j

WalkthrouShe on the BSEP Simulator -

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operator Performance Evaluation on BSEP Simulator 1

Human Factors Review per NUREG-0700 and Standard Review Plan 18.2 t

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l' SPDS DISPLAY WOREING OROL*P l i NED PROJECT  !

MANAGEMENT I

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SPDS DISPLAY

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l LEADER SAM STRICELAND ss ses 1 1 1 I SITE l OPERATIONS SPLS I HUMAN

• PROJECT l COORDINATOR DISPLAY l FACTORS COORDINATOR ENGINEER SPECIALISTS M WILLIAMS l K HORN M BECK l D BEITH

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AS COMPUTER CONSULTANT SUPPORT  !

H EI.aEL s HOLDS SRO LICENSE as MEMBER CRDR HEDAT

EOP DEVELOPMENT

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t DISPLAY DEVELOPMENT P30GRAN I) REVIEW OF GE REV 2 BASED DISPLAYS j

2) NEW DISPLAYS PROPOSED

3) PROPOSED FUNCTIONAL REGUIREMENTS DOCUMENTS WRITTEN

4) MULTI-DISCIPLINARY WORKING GROUP FORMED FOR DISPLAY DEVELOPMENT

6) DRAFT SPECIFICATION WITH FUNCTIONAL REQUIREMENTS

6) WORKING DISPLAYS DEVELOPED

7) REVIEW OF DISPLAYS ,

A. WORKING OROUP B. OPS WALK-THROUGH ON SIMULATOR 6

8) EOP/SPDS OPERATOR TRAINING AND REVIEW .

A. HF OBSERVATIONS IN EIMULATOR B. DISPLAYS REVISED FOR HF CONCERNS i

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MODIFICATION OF THE GE SPDS DISPLAYS 4 s ADDED ALL REV 4 BASED EOP CONTROL PARAMETERS REV 2 OENERIC DISPLAY INFORMATION VS REV 4 BASED EOP LIMITS PLANT STATUS MATRIX ADDED TO ALL DISPLAYS ,

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• PLANT STATUS MATRIX DIGITAL VALUES FOR PC AND RPV PARAMETERS WHITE a

- PARAMETER DESCRIPTIONS AND MESSAGES FOR ALARM BOXES COLOR CODED FOR LIMIT STATUS

- GROUPED LIKE EOPs

- SAME LOCATION ON ALL DISPLAYS a ORGANIZED LIKE EOPs ,

a COLOR CODING GREEN / RED USED FOR OPEN/ CLOSED FOR VALVES

- GREEN, YELLOW AND RED USED FOR PARAMETER LIMIT STATUS

- CYAN BACKGROUND PATTERN USED FOR NOT VALIDATED PARAMETER STATUS ", -

a CLUTTER REDUCTION

- LIMIT BOXES ON CRITICAL PLANT VARIABLES DISPLAY WERE l

REMOVED

- THICK BOXES AND THEIR COLOR CODING WERE REMOVED

! (BOX BORDERS CHANGED SHAPE AND COLOR) i TREND PLOTS

- MADE LARGER VALUE BAR COLOR CODED FOR LIMIT STATUS

- LIMIT TAGS MODIFIED 1

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PROPOSED FUNCTIONAL REQUIREMENTS ,

o CUSTOM DISPLAYS WHICH SUPPORT BSEP EOPs

- DISPLAY AND STATUS ALL 50P CONTROL PARAMETERS

- PERFORM DIFFICULT CALCULATIONS FOR OPERATOR USE THE EXISTING GE DATA BASE WITH AS LITTLE CHANGE AS POSSIBLE ,

s REVIEW BY GROUP COMPOSED OF OPERATIONS, HUMAN FACTORS, COMPUTER SUPPORT, AND GE AS CONSULTANT k

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1 REV. I VS REV. 4 PROCEDURES i

REV 2 BASED PROCEDURES (GE DISPLAYS)

RPV CONTROL PC CONTROL s LEVEL s SUPP POOL LEVEL

PRESSURE e PRESSURE e POWER (REACTIVITY)  : DW TEMPERATURE

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EOP-03 EOP-04 SC CONTROL RAD RELEASE

AREA TEMPERATURE s OFFSITE RELEASE

AREA DIFF TEMPERATURE - MS RAD e AREA RADIATION - OFFGAS RAD ,

! a WATER LEVEL (FLOODING) - SW RAD L s PRESSURE - TB RAD RELEASE s HVAC RAD RELEASE RATE - RB RAD RELEASE

+ - MAIN STACK RAD

- DOSE PROJECTION OTHER DIFFERENCES

METHODS OF CONTROL l  : NEW AND DIFFERENT LIMITS l

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FIGURE 3.3 SPDS DISPLAY GROUPING  ;

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CONTROL CONTROL-y _

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1' FIGURE 3.4 l' PLANT STATUS MATRIX l

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4 ATTACHMENT 4 LICENSEE PRESENTATION ON SPDS OPERATING EXPERIENCE e

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l OPERAT2N0- EXPER1ENCES 1

1 o SOFTWARE PROBLEMS o DATABASE PROBLEMS

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o EQUIPMENT PROBLEMS o ADDITIONAL TRAINING NEEDS 6 o EOP UTILIZATION +

o HISTORICAL DATA UTILIZATION l

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80FTW&RE FR0DLEMS j 1

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o SPTMS DATA NOT ALWAYS AVAILABLE 1

I o 5500 DATA ACQUISITION MODULE DOWHLOAD PROBLEMS o SYSTEM QUE MANAGER PROBLEMS e

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D&T&5&SE PRO 3 LEMS I

o ALARM 5ET POINTS CORRECTIONS NEW CHANGES IMPLEMENTATION DELAYS ,

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.. l EXISTING SYSTEM DATABASE CORRECTIONS o VALVE STATUS CORRECTIONS

.. CAC-SV-1227 C & E -

o OUT OF TOLERANCE DISPLAY VALUES

.. PERCENT REACTOR POWER '

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I EQU1PMENT PRO 3LEMS  !

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SPDS TERMINALS LOCKUP AND HAVE FREQUENT MALFUNCTIONS t

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o SLOW SYSTEM RESPONSE TIME TO OPERATOR QUERY .

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o ALL SCREENS NOT AVAILABLE ON ALL SPDS TERMINALS J I

o CDD OPERATIONS o DATA UPDATE PERIOD IS 700 LONG i

o DIESEL OENERATOR D/A MODULE CHANNEL ISOLATION o SPTMS LOCAL / ERFIS SWITCH FAILURE J o 5500 DATA ACQUISITION MODULE PROBLEMS '

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ADDITION &L TR&INING NEED3 l e VT 100 NODE TRAINING 1

.. SOP. / SPDS TRAIN 7NG COMPLETE o CDD OPERAT20NS 1

.., SCREEN CALL-UP AFTER FA LOVER

.. DOUBLE XEY STROKE PROBLEMS  ;

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9 t e- 0 .

EOP UT1L13&T10N e NUMAN ENGINEERED SCREENS o

OPERATOR ACCEPT!.NCE DURING SINULATOR TRAIN!No e

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N18TORICAL DATA UTIL15AT10N  !

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o UNUSUAL EvgNT ANALyggg  ;

o SYSTEM TROUSLEsHooTING -

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- ' PR0PO8ED RE80LUT1ON 2CNEDULE i

ITEM STATUS SCHEDULE 1 l

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l SPTMS D/A SOFTWARE WORKING ADDITIONAL LONG TERM j l

IMPROVEMENTS NEEDED '

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6500 D/A WORKING ADDITIONAL LONG TERM IMPROVEMENTS NEEDED i l

SPTM8 D/A NARDWARE TROUBLE TICKET LONG TERM FIE BY i

WRITTEN ENGINEERING IN 1990 '

QUE MANAGER WORK AROUND UNKNOWN, REQUIRE 8 i

IN PLACE DENERIC FIE i l

l DATA 8ASE CLEANUP WORK IN PROGRESS COMPLETE IN 3 MONTNS ]

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TERMINAL LOCKUP 8 SUDGETING REPLACEMENT COMPLETE '

' SY END OF 1990 ,

TERMINAL MALFUNCTIONS SUDGETING REPLACEMENT COMPLETE f j ,

sY END Or 1990 l SLOW RESPONSE SUDGETING REPLACEMENT COMPLETE f BY END OF 1990  !

t SCREEN AVAILABILITY BUDGETING REPLACEMENT COMPLETE  !

BY END OF 1990 j ODD OPERATIONS BUDGETING REPLACEMENT COMPLETE I EQUIPMENT BY END OF 1990  !

DATA UPDATE PERIOD COMMERCILL UNKNOWN  !

ME00T!ATIONS I t

DIE 8EL CENERATOR 0/A ROOTCAUSE ANALYSIS FIX COMPLETE IN 1990 )

COMPLETE IN 1989 l t

YT-100 MODE TRAINING IN PROGRESS SIX MONTHS TO ONE YEAR  ;

CDD OPERATIONS IN PROGRESS CONTINUAL EFFORT i TRA!N!NG/ -

FAMILIAR 11ATION

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