ML17157A231
| ML17157A231 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 06/27/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17157A230 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 NUDOCS 9007060248 | |
| Download: ML17157A231 (4) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 ENCLOSURE SAFETY-EVALUATION BY THE. OFFICE-OF NUCLEAR REACTOR. REGULATION SUPPLEMENTAL-SAFETY-EVALUATIONFOR. SAFETY-PARAMETER-DISPLAY SYSTEM SUS UEHANNA STEAM-ELECTRIC. STATION UNITS.l. AND = 2 DOCKET NOS..50-387-AND.50-388
- 1. 0 INTRODUCTION All holders of operating licenses issued by the Nuclear Regulatory Commission (NRC) and applicants for an operating license must provide a Safety Parameter Display System (SPDS) in the control room of their plant.
NRC-approved requirements for the SPDS are defined in Supplement l to NUREG-0737 (Reference l).
The purpose of the SPDS is to provide control room operators with a concise display of critical plant variables to aid in rapid and reliable determination of plant safety status.
Supplement l to NUREG-0737 requires licensees and applicants to prepare a written Safety Analysis describing the basis for determining that parameters used by the SPDS are sufficient to assess the status of specified critical safety functions over a wide range of events.
The symptoms of severe accidents are included.
Licensees and applicants shall also prepare an Imple-mentation Plan for the SPDS which contains schedules for design, development, installation, and full operation of the SPDS as well as f'r design verification and validation.
The Safety Analysis and the Implementation Plan are to be submitted to the NRC for staff review.
Results of the staff review are to be published in a Safety Evaluation Report (SER).
There are eight requirements which the SPDS should satisfy.
They are, with Supplement l to NUREG-0737 references in parentheses, as follows:
l.
Concise display of critical plant variables to aid control room operators in determining the safety status of the plant (4.la).
2.
Location convenient to control room operators (4.lb).
3.
Continuous display of information from which plant safety. status can be assessed (4.lb).
4.
Aid operators in rapid, reliable, determination of plant safety status (4.la and 4.lb).
5.
Suitable isolation from electrical or electronic interference with equipment and sensors that are in use for safety systems (4.1c).
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6.
7.
Incorporation of accepted human-factors principles (4.1e).
Parameters selected to provide, as a minimum, information about reactivity control, reactor core cooling and heat removal from the primary system, reactor coolant system integrity, radioactivity control and containment conditions (4.1f).
8.
Implementation of procedures and operator training leading to timely and correct safety status assessment both with and without the SPDS (4.1c).
NRC staff review was directed at:
1.
Confirming the adequacy of the parameters selected for assessment of critical safety functions.
2.
Confirming that means to assure the validity of displayed data are provided.
3.
Confirming that the licensee or applicant has committed to a human factors program to ensure that displayed information can be readily perceived and comprehended so as not to mislead the operator.
II If, based on its review, the staff identifies a serious safety question or seriously inadequate
- analysis, the Director of the Office of Nuclear Reactor Regulation may request or direct the licensee to cease implementation.
On September 3G, 1983, Pennsylvania Power and Light Company submitted a Safety Analysis Report (SARReference
- 2) for the Susquehanna Steam Electric Station Units 1 and 2 (Susquehanna)'SPDS.
The staff reviewed the licensee's submittal and responded in an SER (Reference
- 3) dated June 2, 1984.
In that SER, several human factors issues were identified as confirmatory issues.
The licensee responded to the confirmatory items by letter dated August 8, 1984 (Reference 4).
An SSER (Reference
- 5) dated February 19, 1985 was sent to the licensee indicating that the licensee's responses to the confirmatory issues were acceptable with one exception.
The remaining issue concerned the use of unconventional color-coding and was relegated to the Detailed Control Room Design Review (DCRDR) for resolution.
This issue was closed as documented in the DCRDR Technical Evaluation Report (Reference
- 6) da'ted November 11, 1986.
In November 1985, the staff and its contractors from Science Applications International Corporation and Comex Corporation conducted an onsite review of the Susquehanna SPDS.
The review team found that the Susquehanna SPDS was
,very effective in providing a continuous, concise display from which operators can rapidly and reliably assess the status of critical safety functions.
,Lastly, the staff issued an SER dated August 29, 1986 (Reference
- 7) concluding that the isolation devices for Susquehanna are acceptable for interfacing the SPDS with safety-related equipment.
- 2. 0 EVALUATION The staff evaluation of the Susquehanna SPDS was consistent with Section 18.2, Revision 0, of the Standard Review Plan, NUREG-0800 (Reference 8).
Because the review team found that the Susquehanna SPDS was fully satisfactory, the licensee was not required to respond to Generic Letter No. 89-06 (Reference 9).
Based on the results of the staff review of the licensee's submittals and the review team finding, the staff again concludes that the licensee has met all of the eight requirements of Supplement 1 to NUREG-0737.
3.0 CONCLUSION
The staff concludes that all requirements for the SPDS at Susquehanna have been satisfactorily met by the licensee.
Therefore, NRC staff review and licensee implementation of the SPDS are considered complete.
Dated:
4
4.0 REFERENCES
1.
NVREG-0737, Supplement 1, "Clarification of TMI Action Plan Requirements-Requirements for Emergency
Response
Capability (Generic Letter No.
82-33)," December 17, 1982.
2.
Letter, from N.
W. Curtis (PPKL) to A. Schwencer (NRC), "Susquehanna Steam Electric Station, SPDS Safety Evaluation,"
September 30, 1983.
3.
Letter, from A. Schwencer (NRC) to N. W. Curtis (PPLL), "SER Regarding Staff Review of Safety Analysis Report for Susquehanna Units 1 and 2,"
June 2, 1984.
4.
Letter, from N.
W. Curtis (PPKL) to A. Schwencer (NRC), "Susquehanna Steam Electric Station SPDS SER Information Request,"
August 8, 1984.
5.
Letter, from A. Schwencer (NRC) to N.
W. Curtis (PP&L), "SER Regarding Human Factors Confirmatory Items for Susquehanna Steam Electric Station Units 1 and 2 Concerning the SPDS," February 19, 1985.
6.
Letter.
R. T. Liner (SAIC) to S.
S. Bajwa (NRC), "Detailed Control Room Design Review Evaluation,"
November 11, 1986.
7.
Memorandum, from F.
Rosa (NRC) to M. Srinivasan (NRC), "Susquehanna-Review of Isolation Devices that Interface with SPDS," August 29, 1986.
8.
NUREG-0800, "Standard Review Plan of Safety Analysis Reports for Nuclear Power Plants,"
Section 18.2, Safety Parameter Display System, November 1984.
9.
Generic Letter No. 89-06, "Task Action Plan Item I.D.2 - Safety Parameter Display System - 10 CFR Part 50.54(f) - (Gener ic Letter No. 89-06),"
April 12, 1989.