ML20235E548
| ML20235E548 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/16/1987 |
| From: | Boyle M Office of Nuclear Reactor Regulation |
| To: | Mroczka E NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| NUDOCS 8709280162 | |
| Download: ML20235E548 (6) | |
Text
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p Docket No.:
50-245 September 16, 1987 Mr. Edward J. Mroczka, Senior Vice President Nuclear Engineering and Operations Northeast Nuclear Energy Company P.O. Box 270 Hartford, Connecticut 06141-0270
Dear Mr. Mroczka:
SUBJECT:
SAFETY PARAMETER DISPLAY SYSTEM-REQUEST FOR ADDITIONAL INFORMATION RE:
Millstone Nuclear Power Station, Unit No. 1 The staff has reviewed the April 9,1987, submittal for Millstone Unit No. I concerning the Safety Parameter Display System and concluded that insufficient information was provided to complete its review. The information needed by the staff to continue its review is described in the enclosure to this letter.
Please provide the response to this request in the next 60 days so that we might complete our review in accordance with the proposed ISAP integrated implementation schedule.
The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respor. dents; therefore, OMB clearance is not required under P.L.96-511.
original signed by Michael L. Boyle, Project Manager Integrated Safety Assessment Project Directorate Division of Reactor Projects III, IV, V and Special Projects
Enclosure:
As stated cc:
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Concurrence l
09//6/87 09ff,/87 J
ta Mr. Edward J. Mroczka Mills' tone. Nuclear Power Station Northeast Nuclear Energy Company Unit No. 1 cc:
Gerald Garfield, Esquire Kevin McCarthy, Director Day, Berry & Howard Radiation Control Unit Counselors at Law Department of Environmental City. Place Protection Hartford, Connecticut 06103-3499 State Office Building Hartford, Connecticut 06106 Wayne D. haberg-Richard M. Kacich, Manager Vice President, Nuclear Operations Generation Facilities Licensing Northeast Utilities Service Company Northeast Utilities Service Company
. Post Office Box 270 Post Office Box 270 Hartford, Connecticut 06141-0270 Hartford, Connecticut 06141-0270 Bradford S. Chase, Under Secretary Donald O. Nordquist, Director Energy Division Quality-Services Department Office of. Policy and Management Northeast Utilities Service Department 80 Washington Street Post Office Box 270 Hartford, Connecticut' 06160 Hartford, Connecticut 06141-0270 Regional Administrator, Region I J. Stetz, Pla.. Superintendent U.S. Nuclear Regulatory Comission Millstone, Unit 1 631 Park Avenue Northeast Nuclear Energy. Company King of Prussia, Pennsylvania 19406 P. O. Box 128 Waterford, Connecticut 06385 Northeast Nuclear Energy ~ Company ATTN:
Superintendent Millstone Nuclear Power Station P. O. Box 128 Waterford, Connecticut 06385 Resident Inspector c/o U.S. NRC Millstone Nuclear Power Station P. O. Box 811 Niantic, Connecticut 06357 First Selectman of the Town of Waterford Hall of Records 200 Boston Post Road Waterford, Connecticut 06385
r REQUEST FOR ADDITIONAL INFORMATION CONCERNING THE MILLSTONE NUCLEAR POWER STATION, UNIT 1 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 (Reference y).
In the Regional Workshops on Generic Letter 82-33 held during March 1983, the NRC discussed these l
requirements and the staff's review of the SPDS.
Prompt implementation of the SFDS in operating reactors is 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 inade directed at (a)quate.
The NRC staff review to accomplish this will be 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 committ6d 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 (d) confirming that the SPDS will be suitably isolated from electricci 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 NRR may require or direct the licensee to cease implementation.
The staff reviewed the SPDS safety analysis and implementation plan provided by Northeast Utilities (Reference 2). The staff was unable to complete its evaluation because of insufficient information. The following additional information is required to continue and complete the SPDS evaluation:
PARAMETER SELECTION The staff's review of the variables selected for display on the Millstone 1 SPDS identified the following concern:
In addressing the Reactivity Control CSF, the licensee lists two parameters, Reactor Power and Scram Demand.
It is not clear whether the Reactor Power parameter is representative of the full range of reactor power (0 - 100% +).
It is implied that the input to thi., parameter is an Average Power Range Monitor (APRM) signal. The licensee should be aware that the staff believes that parameters chosen to represent the Reactivity CSF should provide the capability to monitor power levels down to and in:luding zero power.
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Acceptable parameters covering this range are APRM and. Source Range Monitors (SRM) or if.SRM data is infeasible, APRM and control rod' position. Other alternate parameters providing the same information will be considered.
The licensee should provide a commitment to' add parameters to the SPDS that will allow users to monitor Reactivity Control over the full range of reactor power levels down to and including zero power, or provide further explanation regarding how the presently selected parameters accomplish ~ the same function.
-DATA VALIDATION The staff evaluated the licensee's design to determine that means are provided in the. display. system to assure that the data displayed are valid. The following concern was identified:
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In its submittal the licensee describes the data range-checking and validation, limit checking, and status evaluation functions of the Real-Time Analysis and Display Function of the SPDS.. It could not be determined if all j
signals are subjected.to this process. Specific status of the validation process is available to the operator via the' validation status displays for the'following parameters:
RPV Level, Pressure, and Temperature Reactor Power Drywell Pressure Drywell/Suppres.sion Chamber Temperature Suppression Pool Level and Temperature Reactor Core Flow It appears that containment radiation, main stack radiation, main steamline radiation, main steamline isolation valve (MSIV) status, safety relief valve (SRV) status, and hydrogen / oxygen concentration are not subjected to any validation process.
The licensee should submit further information detailing the method of1 validation used for each of the selected safety parameters, or should justify not providing data validation.
Emphasis should be placed on the parameters identified in the paragraphs above because they constitute the set of parameters that would be minimally acceptable to the staff.
HUMAN FACTORS PROGRAM The staff notes that the licensee has committed to meeting the requirements of Supplement 1 to NUREG-0737. However, a commitment to providing continuous display was omitted. The staff position is that the SPDS parameters must be continuously displayed (e.g.-dedicated display) or an alerting mechanism provided so that the operator is aware of changes in critical safety function status and can easily access the parameter (s) that are affecting the status of combined with function key access to detailed data) played status lights the critical safety function (e.g. continuously dis
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< The licensee should provide a description of the Millstone l' SPDS design in enough detail that the staff is able to confinn that the minimum information
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sufficient-to monitor plant safety status is continuously displayed, in accordance with the staff position described above.
IS01.ATION DEVICES For each type of device used'to accomplish electrical isolation, a.
describe the specific testing performed to demonstrate that the device is acceptable for its application (s). This description should include elementary diagrams where necessary to indicate the test configuration and how the maximum credible faults were applied to the
- devices, b.
Provide 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, i
Provide data to verify that the maximum credible fault was applied to c.
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.
Provide a commitment that the isolation devices comply with the e.
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.
Provide information to verify that the Class IE isolation is powered g.
from a Class 1E source.
UNREVIEWED SAFETY QUESTIONS Provide a cummary of the conclusions drawn by the Plant Operations Review Committee and the Nu lear Review Board regarding the Millstone 1 SPDS.
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REFERENCES 1.
U.S. Nuclear Regulatory Commission, " Clarification of TMI Action Plan Requirements, 'viequirements For Emergency Response Capability,"
U.S. NRC Report NUREG-0737, Supplement 1, January 1983.
l 2.
Letter form E.J. Mroczka (Northeast Nuclear Energy Company) to NRC Document Control Desk, with attachments, dated April 9, 1987.
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