ML20209G300

Submits Review & Justification for Changes in Final SPDS Displays & Parameters from Displays & Parameters Initially Identified in Original Sar.Draft Rept of Emergency Response Computer Sys Availability Study Encl

ML20209G300
Person / Time
Site:	Prairie Island
Issue date:	01/26/1987
From:	Musolf D NORTHERN STATES POWER CO.
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20209G304	List: ML20209G300 ML20209G344
References
NUDOCS 8702050332
Download: ML20209G300 (4)

Text

/6 E ff Northem States Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401 Telephone (612) 330-5500 January 26, 1987 Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Review of SPDS As-built Against the Approved Safety Analysis Report As requested herein is a review and justification for changes in the final SPDS displays and parameters from the displays and parameters initially identif*ed in the original Safety Analysis Report. This review was requested by Mr. Dominic DiIanni, Project Manager, Project Directorate #1, Division of PWR Licensing A.

REFERENCES:

1. SPDS Safety Analysis Report NSP-NRC letter dated April 10, 1984 l

l 2. Request for Additional Information NRC-NSP letter dated February 14, 1985

3. Additional Information Related to SPDS NSP-NRC letter dated April 17, 1985

4. SPDS Final Evaluation NRC-NSP letter dated July 15, 1985

5. SPDS Exit Meeting with James Hard, NRC Prairie Island Senior Resident Inspector, on December 31, 1986 The changes are as follows:

1. The Accident Identification Display System (AIDS) has not been imple-mented on the Emergency Response Computer System.

Justification: As identified in Section 2.2.2 of the Safety Analysis Report, AIDS is outside the scope of the SPDS Requirements.

8702050332 870126 PDR ADOCK 05000282 P PDR tti

Office of Nuclear Reactor Regulation Page 2 January 26, 1987

2. Section 2.3.4 Control Room Location and Figure 2-1 indicate that a CRT will be located on the lead operators desk. This CRT was not incor-ported in the final design.

Justification: This CRT location was intended to be used by the STA during an emergency situation. Observations of the STA during simulator exercises shows that the STA plays a much more active role with the operators and would be using either the primary display CRT or the display located on the operator console.

3. Section 2.3.4 Control Room Locat'.on states that the critical top level data is readable to a distance af 15 feet. The data cannot be read from a distance of 15 feet.

Justification: A human factors review (Design Review 14) was performed in accordance with the SPDS Verification and. Validation Plan. The displays were evaluated in accordance with NUREG 0737 Supplement 1 Section 4.1b guidelines concerning location to the users and readability.

The review showed compliance with human factors requirements for both location and readability.

4. Attachment 1, SPDS Monitored and Displayed Parameters The following parameters are not displayed parameters on the SPDS system but were in the original designs as indicated in the Safety Analysis Report.

a. Intermediate Range Startup Rate Justification: The startup rate is a calculated value and is used in the Subcriticality Status Trees to identify a possible ATWS event.

The SPDS Nuclear Instrumentation Trend displays the log power range. This parameter may be trended to indicate a startup rate.

The CRDR Task Analysis for Emergency Operating Procedures shows that the startup rate is used in the Response to Nuclear Power Generation /ATWS Function Restoration Procedure as an action point to ensure the reactor is subcritical. This information can be readily obtained by reviewing the Subcriticality Status Trees.

b. Reactor Trip Status Justification: The trip status is not directly displayed on the SPDS displays. If a reactor trip occurred, it would be indicated on the alarm screen CRT located on the control room operators work station and would also be indicated on the SPDS displays as " Auto Event". The first out reactor trip annunciator would also alert the operator of a reactor trip.

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' Office of Nuclear-Reactor Regulation Page 3 January 26, 1987

c. Auxiliary Feedwater Flow Justification: The Auxiliary Feedwater Flow is used in the Heat Sink Status Tree to identify insufficient flow to the steam gen-erators. It is used strictly as an action point to identify the proper procedure to be used if Steam Generator Level is not proper and Auxiliary Feedwater Flow is less than 200 gpm. This action point information can be readily obtained by reviewing the Heat Sink Status Tree.

5. Attachment 2 SPDS Parameter Ranges The following ranges / limits are different than those described in the Safety Analysis Report.

a. Reactor Power Average Power Range changed from 0-125% to 0-120%.

Justification: This parameter is an input from the Reg. Guide 1.97 Gamma Metrics Nuclear Instrumentation. The Gamma Metrics instru-mentation was in the design stage when the Safety Analysis Report was written. The final design for Gamma Metrics has an Average Power Range of 0-120%, thus the SPDS range is 0-120%.

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b. Containment Radiation ranged changed from 1-10 R/ Hour to 1-10 R/ Hour.

Justification: The range was changed to be consistent with the monitor's total range. The low range was eliminated. The action point for critical safety function entry into the Response to High Containment Radiation Level Function Restoration Procedure is containrent radiation level > 200 R/Hr. This level is well above the low range monitor's high limit.

c. Containment Sump Level Units were changed from 0-144 inches to 0-12 feet.

Justification: The control board instruments are calibrated and displayed in feet, not inches. All procedure action points are also indicated in feet,

d. Main Stack Effluent units are cpm and not mR/Hr.

Justification: This was an error in our Safety Analysis Report.

The units for the input instruments are cpm.

Office of Nuclear Reactor Regulation Page 4 January 26, 1987 These changes as described and justified do not change the intent of the SPDS design as described in the Safety Analysis Report nor do they change the conclusions addressed in the Preliminary 10CFR50.59 Safety Evaluation in-cluded in the Safety Analysis Report.

Also attached is a copy of the draft report of the Availability Study for the SPDS. This study is in the process of being reviewed by NSP.

Please contact us if you require additional information related to your request.

David Musolf Manager-Nuclear Support Se ices cc: Regional Administrator-III NRR Project Manager, NRC Resident Inspector, NRC G Charnoff