ML17328A700

Responds to NRC 900406 Ltr Re Inadequacies of Spds,Per Audit on 900221-22.Corrective Actions:Software Mod Will Prevent User from Accessing Displays Other than Iconic Displays from SPDS Dedicated Terminal

ML17328A700
Person / Time
Site:	Cook
Issue date:	05/09/1990
From:	Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-0737, RTR-NUREG-737 AEP:NRC:0773AH, AEP:NRC:773AH, NUDOCS 9005150150
Download: ML17328A700 (6)

Text

ACCELERATED . >STjUBUTION DEMON TION SYSTjM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9005150150 DOC.DATE: 90/05/09 NOTARIZED: NO DOCKET 45 FACIL:50-315 Donald C. Cook Nuclear Power Plant, Unit 1, Indiana 6 05000315 50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana & 05000316 AUTH. NAME AUTHOR AFFILIATION ALEXICH,M.P. Indiana Michigan Power Co. (formerly Indiana 6 Michigan Ele RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Responds to NRC 900406 on 900221-22.

ltr re inadequacies of SPDS,per audit DISTRIBUTION CODE: A003D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: OR/Licensing Submittal: Suppl 1 to NUREG-0737(Ge eric Ltr 2-33)

NOTES'ECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-1 LA 1 1 PD3-1 PD 7 7 GIITTER,J. 1 1 INTERNAL: NRR HFB11 1 1 OC/LFMB 1 0 ILE 01 1 1 RES/DSIR/EIB 1 1 EXTERNAL: LPDR 1 1 NRC PDR 1 1 NSIC 1 1 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASIEi CONTACT THE.DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISHUBUTION LISIS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 16 ENCL 15

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Indiana Michigan Power Company

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P.O. Box 16631 Coiumbus, OH 43216 AEP:NRC:0773AH Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 License Nos. DPR-58 and DPR-74 RESPONSE TO NRC AUDIT OF COOK NUCLEAR PLANT SAFETY PARAMETER DISPLAY SYSTEM (SPDS) INADEQUACIES U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Attn: T. E. Murley May 9, 1990

Dear Dr. Murley:

This letter responds to your April 6, 1990, letter and provides the corrective actions we have taken, or plan to take, to correct the findings regarding the on-site audit of the SPDS.

On February 21-22, 1990, the NRC audited the SPDS using the criteria given in NUREG-0737 Supplement 1. The NRC audit team determined that the Cook Nuclear Plant SPDS for both Units 1 and 2 met five of the eight requirements of NUREG-0737, Supplement 1. We agreed to respond to the NRC audit team's findings regarding the three requirements not met. The three findings and our responses are as follows.

SPDS Is Not Continuous o The SPDS monitor does not provide any visual or audible cues to alert operators to a change in status for the five plant-specific safety functions on displays other than the two, top level iconic displays.

~Res ense The SPDS does not provide a continuous display on the Technical Support Center (TSC) computer terminals located in each control room. Operators do have the ability to access lower level displays on these terminals for more detailed information.

The immediate solution to correct this inadequacy will be a software modification. The modification will limit the keyboard functions available on the control room TSC computer terminal and 9005i50i50 900509 PDR ADOCK 050003i5 P PDC o8

Dr. T. E. Murley AEP:NRC:0773AH provide a dedicated display of the SPDS iconic. The software modification will prevent a user from accessing any displays other than the iconic displays from the SPDS-dedicated terminal.

A long-range solution is also being investigated that would provide the operator with a signal of any adverse changes to the iconic displays'rocess variables, which would permit a user to access other displays on the SPDS-dedicated terminal. If this alternative solution is found to be practical it will be implemented.

SPDS Is Not Desi ned To Incor orate Acce ted Human Factors Princi les o Numerical values for the high and low level alarm setpoints on each safety function "spoke" of the narrow- and wide-range iconic displays were not identified. Some alarm setpoints were reactor trip values and others were either anticipatory values or system capacity values. In addition, the Unit 2 SPDS low steam generator low level reactor trip setpoint, 21%, was incorrectly set at the Unit 1 value of 17%. In another case, the Unit 1 SPDS high Tavg alarm setpoint on the SPDS had not been changed to reflect the amended Technical Specification value.

R~es ense There is not enough physical space to display all of the high and low limit values on the narrow- and wide-range iconic From a human factors standpoint, the fact that the displays'pokes.

discrete high and low limits for each spoke are not displayed does not detract from the functionality of the iconic display concept. The iconic displays provide a graphical representation of the status of specific process variables. It is not necessary for a spoke's high and low limits to be discretely displayed for the functionality and usefulness of the iconic displays to be fully realized. It would actually detract from the primary design concept of the iconic displays, which is rapid recognition of a critical safety function problem. Therefore, the high and low limits will not be added to the spokes of the iconic displays. The apparent disparity in the selection of reactor trip values, system capacity values, or anticipatory values will be rectified.

For the narrow-range iconic display, which is primarily used during normal power operations, the high and low limits for each spoke will be based primarily on reactor trip setpoints or engineered safety system actuation setpoints, if they are If applicable to a variable on the narrow-range iconic display.

there are no reactor trip or engineered safety system actuation setpoints for a particular variable, such as the RCS average temperature, Technical Specification limiting condition for operation values will be used for a spoke's high and low limit values. If there are no reactor trip setpoints, engineered

Dr. T. E. Murley AEP:NRC:0773AH safety system actuation setpoints, or Technical Specification limiting condition for operation values for a spoke's high and low limits, then annunciator setpoints will be used. Finally, none of the above criteria are applicable to a specific process if variable on the narrow-range iconic display, then system capacities or other appropriate values will be used, as is the case for net charging flow and power mismatch.

For the wide-range iconic display, which is used primarily after a reactor trip has occurred related to an accident condition, the high and low limits for each spoke will be based primarily on emergency operating procedure status tree values, as applicable, or alternatively on physically meaningful or appropriate values.

The narrow-range steam generator level low limit on the Unit 2 narrow-range iconic display was changed to the correct reactor trip setpoint. The RCS average temperature high limit on the Unit 1 narrow-range iconic display had not been changed to reflect a recent amendment to the Technical Specification limiting condition for operation value. The Unit 1 primary syst: em recently was modified for reduced temperature and pressure operation. Changes resulting from this modification were evaluated and, as applicable, were implemented on the TSC computer system. The high limit value has been changed to the correct value corresponding to the amended Technical Specification limiting condition for operation requirement.

0 erator Trainin With SPDS Was Not Satisfactor o Control room operators did not know, in all instances, what the numerical values were for the high and low level alarm setpoints for each safety function on the SPDS narrow- and wide-range iconic displays.

R~es ense The numerical values for the high and low alarm limits for each spoke on the narrow- and wide-range iconic display will be provided to the Cook Nuclear Plant operators as a technical data reference. The general philosophy on how these values were selected will be incorporated into the operator requalification training beginning in the third quarter of 1990.

Dr. T. ED Murley AEP:NRC'0773AH This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.

Sincerely, M. P. Alexich Vice President ldp cc: D. H. Williams, Jr.

AD A. Blind - Bridgman R. C. Callen G. Charnoff A. B. Davis - Region III NRC Resident Inspector - Bridgman NFEM Section Chief