ML15264A257
| ML15264A257 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/14/1984 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Tucker H DUKE POWER CO. |
| References | |
| NUDOCS 8409280512 | |
| Download: ML15264A257 (6) | |
Text
iockets Nos. 50-269 i
u 50-270 e tFi 50-287 NRC & L PDRs Gray Files Reading Files DEisenhut Mr. Hal B. Tucker ACRS 10 Vice President - Steam Production EJordan Duke Power Company JNGrace P. 0. Box 33189 GDick 422 South Church Street GWLapinsky Charlotte, North Carolina 28242 EBlackwood HOrnstein
Dear Mr. Tucker:
RIngram HNicolaras
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION:
SAFETY PARAMETER DISPLAY SYSTEM RE:
Oconee Nuclear Station, Units 1, 2 and 3 We have reviewed your February 16, 1984 submittal on the Safety Parameter Display System (SPDS) and have concluded that additional information is necessary to complete our evaluation. Please respond within 30 days of receipt of this letter to the enclosed request for additional information.
This request for information affects fewer than ten respondents, therefore, OMB clearance is not required under P. L.96-511.
Sincerely, John F. Stolz, Chief Operating Reactors Branch No. 4 Division of.Licensing
Enclosure:
As stated cc w/enclosure:
See next page OR 4DL 0
L aras;ef JFS 09/
/84 09/
4 8409280512 840914 PDR ADOCK 5000269 P
PDR_
REUEST FOR ADDITIONAL INFORMATION CONCERNING THE OCONNE 1, 2, AND 3 SAFETY PARAMETER DISPLAY SYSTEM Each operating reactor shall be provided with a Safety Parameter Display System (SPDS).
The Commission approved requirements for an SPDS are defined in NUREG-0737, Supplement 1. In the Regional Workshops on Generic Letter 82-33 held during March 1983, the NRC discussed these requirements and the staff's review of the SPDS.
Prompt implementation of the SPDS in operating reactors is a design goal of prime importance.
The staff's review of SPDS documentation for operating reactors called for in NUREG-0737, Supplement 1 is designed to avoid delays resulting from the time required for NRC staff review. The NRC staff will not review operating reactor SPDS designs for compliance with the requirements of Supplement 1 of NUREG-0737 prior to implementation unless a pre-implementation review has been specifically requested by licensees.
The licensee's Safety Analysis and SPDS Implementation Plan will be reviewed by the NRC staff only to determine if a serious safety question is posed or if the analysis is seriously inadequate. The NRC staff review to accomplish this will be directed at (a) confirming the adequacy of the parameters selected to be displayed to detect critical safety functions, (b) confirming that means are provided to assure that the data displayed are valid, (c) confirming that the licensee has committed to a human factors program to ensure that the displayed information can be readily perceived and comprehended so as not to mislead the operator, and (a) confirming that the
SPDS will be suitably isolated from electrical and electronic interference with equipment and sensors that are used in safety systems. If based on this review, the staff identifies a serious safety question or seriously inadequate analysis, the Director of Inspection and Enforcement (.IE) or the Director of Nuclear Reactor Regulation (NRR) may require or direct the licensee to cease implementation.
The staff reviewed the SPDS Safety Analysis and Implementation Plan provided by Duke Power Company (Reference 1).
The staff was unable to complete its evaluation because of insufficient information. The following additional information is required to continue and complete the SPUS evaluation:
ISOLATION DEVICES
- a.
For each type of device used to accomplish electrical isolation, describe the specific testing performed to demonstrate that the device is acceptable for its application(s).
This description should include elementary diagrams when necessary to indicate the test configuration and how the maximum credible faults were applied to the devices.
- b. Data to verify that the maximum credible faults applied during the test were the maximum voltage/current to which the device could be exposed, and define how the maximum voltage/current was determined.
-3
- c. Data to verify that the maximum credible fault was applied to the output of the device in the transverse mode (between signal and return) and other faults were considered (i.e., open and short circuits).
- d. Define the pass/fail acceptance criteria for each type of device.
- e.
Provide a commitment that the isolation devices comply with the environmental qualifications (10 CFR 50.49) and with the seismic qualifications which were the basis for plant licensing.
- f. Provide a description of the measures taken to protect the safety systems from electrical interference (i.e.,
Electrostatic Coupling, EMI, Common Mode and Crosstalk) that may be generated by the SPDS.
HUMAN FACTORS PROGRAM Provide a description of the display system, its human factored design, and the methods used and results from a human factors program to ensure that the displayed information can be readily perceived and comprehended so as not to mislead the operator.
-4 DATA VALIDATION Describe the specific methods used to validate data displayed in the SPDS. Also describe how invalid data is defined to the operator.
IMPLEMENTATION PLAN Provide a schedule for full implementation of the SPDS including hardware, software, operator training, procedures and users manuals.
REFERENCE Letter from Duke Power Company to NRC dated February 16, 1984, forwarding Revision 3 of the response to Supplement 1 to NUREG-0737.