ML17055B641
| ML17055B641 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 05/08/1986 |
| From: | Zwolinski J Office of Nuclear Reactor Regulation |
| To: | Mangan C NIAGARA MOHAWK POWER CORP. |
| References | |
| TASK-1.D.1, TASK-TM NUDOCS 8605150571 | |
| Download: ML17055B641 (18) | |
Text
0 May 8, 1986 s
l Docket No. 50-220 Niagara Mohawk Power Corporation ATTN:
Mr. C., V. Mangan Senior Vice President c/o Miss Catherine R. Seibert 300 Erie Boulevard West
- Syracuse, New York 13202
Dear Mr. Mangan:
SUBJECT:
SAFETY EVALUATION - SAFETY PARAMETER DISPLAY SYSTEM (TAC 51259)
Re:
Nine Mile Point Nuclear Station, Unit No.
1 Enclosed is our Safety Evaluation regarding the Nine Mile Point 1 Safety Parameter Display System (SPDS).
We have identified no serious safety question during our review and conclude that implementation of the SPDS may continue.
However, the staff's review is not complete.
You should provide the information identified in Sections 2.2 and 2.4 of the Safety Evaluation in order to assist our review on a schedule to be negotiated with your PM.
The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.
Sincerely, mzoT John A. Zwolinski, Director BWR Project Directorate ¹1 Division of BWR Licensing
Enclosure:
Safety Evaluation cc w/enclosure:
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Mr. C. V. Mangan Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station, Unit No.
1 CC:
Troy B. Conner, Jr
, Esquire Conner 5 Wetterhahn Suite 1050 1747 Pennsylvania
- Avenue, N.
W.
D. C.
20006 Frank R. Church, Supervisor Town of Scriba R.
D.
82
- Oswego, New York 13126 Niagara Mohawk Power Corporation ATTN:
Mr. Thomas Perkins Plant Superintendent Nine Mile Point Nuclear Station Post Office Box 32
- Lycoming, New York 13093 Resident Inspector U. S. Nuclear Regulatory Commission Post Office Box 126
- Lycoming, New York 13093 John W. Keib, Esquire Niagara Mohawk Power Corporation 300 Erie Boulevard West
- Syracuse, New York 13202 Regional Administrator, Region I'.
S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 Mr. Jay Qunkleberger Division of Policy Analysis and Planning New York State Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO TMI ITEM I.D.2 - SAFETY PARAMETER DISPLAY SYSTEM NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NIICLEAR STATION, UNIT 1
1.0 INTRODUCTION
All holders of operating licenses issued by the Nuclear Regulatory Commission (the licensees) and applicants for an operating license (OL) must provide a
Safety Parameter Display Systems (SPDS) in the control room.
The Commission approved requirements for the SPDS are defined in Supplement 1 to NUREG-0737.
The purpose of the SPDS is to provide a concise display of critical plant variables to control room operators to aid them in rapidly and reliably determining the safety status of the plant.
NUREG-0737, Supplement 1,
requires licensees and applicants to prepare a written safety analysis describing the basis on which the selected parameters are sufficient to assess the safety status of each identified function for a wide ranqe of
- events, including symptoms of severe accidents.
Licensees and applicants shall also prepare an Implementation Plan for the SPDS which contains schedules for design, development, installation, and full operation of the SPDS as well as a design Verification and Validation Plan.
The Safety Analysis and the Implementation Plan are to be submitted to the NRC for staff review.
The results from the staff's review are to be published in a Safety Evaluation (SE).
Prompt implementation of the SPDS in operating reactors is a design goal of prime importance.
The'taff's review of SPDS documentation for'perating reactors contained 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 speci,fically 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 (CSF),
(b) confirminq 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 (d) 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 seriousl,v inadequate
- analysis, the Director of IE or the Director of NRR may require or direct the licensee to cease implementation.
I
2.0 EVALUATION Niagara Mohawk Power Corporation (NMPC) submitted several documents in which the Nine Mile Point, Unit No. I (NMP-1)
SPDS analysis and design process were described.
The staff has reviewed these documents, and the results of that review are presented below.
2.1
~222 Il The Nine Mile Point I SPDS is a computer-based data acquisition and display subsystem that is implemented on the plant process computer.
The SPDS was
'designed and developed by NMPC staff using industry guidance.
The display philosophy is conceptually based on the information and format of the Boiling l~'ater Reactors Owners Group (BWROG) Graphic Display System (GDS).
There are two major levels of display.
The top level is a display showing bar-qraphs representing the CSF.
The second level display provides detailed information regarding each of the five CSF.
2.2 Parameter Selection Section
- 4. 1f of Supplement I to NUREG-0737 states that:
"The minimum information to be provided shall be sufficient to provide information to plant operators about:
(i)
Reactivity control (ii)
Reactor core cooling and heat removal from the primary system (iii)
Reactor coolant system integrity (iv)
Radioactivity control (v)
Containment conditions."
For review purposes, these five items have been designated as CSF.
The selection of the NMP-1 SPDS display variables was made by the licensee based on industry guidance provided in NSAC-21 (Ref. 5),
NRC Requlatorv Guide 1.97 (Ref.
- 6) and plant-specific Emergency Procedure Guidelines (EPG).
Extensive supporting discussion was also provided in the licensee's January 3,-
1984 submittal with additional information provided in Reference 4.
The variables selected by the licensee and their relationship to the CSF are summarized in the attached Table 1 (qrouping was made by the licensee).
Insufficient information is provided to evaluate the adequacy of the simulated input used in validation testing.
Specifically, the transient and accident sequence test cases used for performance tests of the SPDS should be'rovided and justified.
Reference to the transients identified in the report "Simulator Evaluation of the Boiling Mater Reactor Owners Group (BWROG) Graphic Display Systems (GDS)" may be acceptable; however, this information should be supplemented to identify tests of the radioactivity control instrumentation.
Also, additional discussion should be provided to cover beyond design'asis conditions.
~
3 The staff has reviewed the licensee's Safety Analysis Report on the NMP-I SPDS.
Based on this review of the licensee's supporting analyses, and the observation that the selected variables appear to be consistent with the BWROG EPG, the staff finds the proposed list of key variables to be generally acceptable.
However, additional discussion of the validation tests should be provided for the staff's confirmatory review to demonstrate the usability of the SPDS.
Also, design flexibilityshould be provided for possible future expansion of the SPDS.
For example, with consideration of the BWROG EPG and with possible amendments to the EPG, other key variables may be identified to assess the safety status of the CSF.
2.3 Dis lay Data Validation The staff reviewed the licensee's submittals to confirm that means have been provided to assure that the data displayed are valid.
The licensee stated in its submittal dated September 18, 1984 (Ref. 3) that the proposed method of data validation consists of comparing at least two independent sensor inputs, using redundancy checks for consistency and out of range checks for reasonableness.
The SPDS utilizes both'nalog and diaital parameters.
Data processing consists of analog data validitv checking, digital filtering of analog data, alert or failed analog sensor identifi-
- cation, alarm limits checking, input error checking, and "chattering" eliminations.
Based on this use of ranqe and status
- checking, as well as physical redundancy, the staff's judgment is that the licensee has provided means to assure that displayed data are valid.
2.4 Human Factors Proqram The staff reviewed the licensee submittals to determine whether a
human factors program had been committed to as part of the SPDS design process.
Reqarding a
human.factors program, the licensee has based the NMP-I SPDS on the GDS developed by the RWROG which included several iterations of human factors review and testing.
In addition, the licensee states (Ref.
- 1) that the SPDS displays will be further reviewed durinq the NMP-I control room design review.
Based on the above statements, the staff concludes that NMPC has committed to a
human factors program in the design and implementation of its SPDS.
However, the staff did note a potential problem in the proposed design regarding the requirement for "continuous display".
It is the staff's position that a continuous display of plant safety status must be provided.
Examples of how this may be accomplished are:
(I) a dedicated display, such as a cathode ray tube continuously displays the minimum set of variables necessary to assess the safety status of the plant; or
(2) a hierarchical display system is used with control room operator-controlled means to quickly and easily access all levels of display formats needed to evaluate the safety status of the plant, and (3) perceptual (audible or visual) cues are provided by the system to alert the control room operator to return to the primary display format while viewing secondary information.
The main concern is that SPDS users are made aware of important chanqes in critical safety-related variables when they occur and that SPDS users can readily obtain information from the SPDS to help them determine the safety status of the plant.
The licensee should provide further clarification regarding how the proposed design will fulfillthe NUREG-0737 Supplement I, requirement for continuous display of plant safety status.
2.5 Electrical and Electronic Isolation The initial Safety Analysis Report submitted by NMPC (Ref. I) did not address the requirement for isolation of the SPDS from equipment and sensors that are used in safety systems to prevent electrical and electronic interference.
In June of 1984, a request was sent for additional information on the isolation devices that are used between safety systems and the SPDS.
The licensee responded on September 18, 1984 (Ref. 3).
The information submitted
. did not address an actual transverse mode test using the maximum credible fault voltage and current.
The submittal discussed the analysis performed by the vendor, Rochester Instrument Systems (RIS) Co., which stated that if the maximum credible fault were applied in the transverse mode across the output terminals, internal component damage would result in the output stage, but that the input side would not he affected.
In telephone conversations on February 27,
- 1985, and March 12, 1985, with plant personnel, the transverse mode test was discussed.
As a result of the discussion, the licensee submitted details of an actual transverse mode test on a typical RIS Co. isolator.
The NMP-I SPDS is driven using input from the plant computer which is a
Honeywell 4400 process management system.
The svstem can be programmed to display any of the plant parameters which input to the system on its color graphics display subsystem.
The computer and its display subset are non-Class 1E equipment.
As such, the system design includes isolation devices at interfaces between safety-related circuits and nonsafety-related circuits in an effort to reduce the potential for the computer system to adversely offset safety-related circuits and/or systems.
The isolators used are RIS Co. Model SG-326 isolated millivolttransmitters.
There are three different versions of the isolators and they differ by the size (value) of the resistor in the output section.
This resistor sets the input/output range of the isolator in millivolts, i.e.,
16-80 mv, 0-160 mv, 0-80 mv.
Otherwise, the internal circuitry is identical for the three isolators.
~
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3.0 The vendor, RIS, conducted the tests on a typical SC-326 isolator.
The isolator passed common mode and dielectric tests, open and short circuits with no adverse effects on the input.
Upon application of the maximum credible fault in the transverse mode to the output, the isolator was damaged as a result of combustion, but maintained isolation:
Damage was limited to the output stage and no perturbations carried through to the input stage or to the circuits connected to the input terminals.
The maximum credible fault was determined to be 120 VAC at 20 amps.
The RIS document "Test Procedure for SC-326 and SC-326W, RIS A-1008-385" describes the production tests for the isolators and also defines the appropriate pass/fail acceptance criteria for conditions pertinent to the NRC's requests.
A certificate of compliance was provided to NMPC for a representative number of the installed isolators.
Based on the staff's review of the documentation submitted by NMPC with respect to the RIS SC-326 analog isolators, the staff concludes that these devices are qualified isolators and are acceptable for interfacing the Class IE safety systems with the SPDS.
The staff also concludes that these devices meet the Commission's.requirements of Section 4 in NUREG-0737, Supplement No. l.
CONCLUSIONS Based on its review of the NMP-I SPDS, the staff finds that:
The variables selected for display are adequate to assess CSF.
If implemented as designed, the SPDS will be suitably isolated from plant safety systems.
The licensee's design provides means to assure that displayed data are valid.
The licensee has. committed to conduct a
human factors engineering program which will allow reasonable assurance that the information provided will be readily perceived and comprehended by its users'..
To date, the staff has not identified any serious safety question or inadequacy in the licensee's analysis and, therefore, finds no.reason to direct the licensee to cease implementation.
However, the staff's review is not complete.
The licensee should provide the information listed in Sections 2.2 and 2.4 in order to assist the staff in completing its review.
The conclbsion that SPDS implementation may continue does not imply that the SPDS meets the requirements of Supplement 1 to NUREG-0737.
That determination can only be made after a post-implementation audit or when the staff has otherwise obtained sufficient information.
Principal Contributor:'. Lapinsky Dated:
May 8, 1986.
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TABLE 1 1.
Reactor Control APRM IRM SRM Control Rod Position 2.
Reactor Core Coolin Reactor Water Level Core Flow I
3.
Reactor Coolant S stem Inte rit Reactor Pressure Drywell Sump Flow Rate Drywell Pressure I
Safety/Relief Valve Position 4.
Containment Integrity Drywell Pressure Torus Pressure Main Steam Isolation Valve Position Suppression Pool Temperature Suppression Pool Level
'rywell Temperature I
Oxygen Concentration 5.
Radioactivi t Control Main Stack Release Rate Containment Activity Off-gas Dose Rate
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4.0 REFERENCES
1.
Letter from C. V. Mangan (NMPC) to D. B. Vassallo (NRC) dated January 3,
1984.
2.
Meeting between NMPC and NRC, February 27, 28, and 29, 1984 at Unit 1 site.
3.
Letter from C.
V. Mangan (NMPC) to D. B. Vassallo (NRC) dated September 18, 1984.
4.
Letter from C. V. Mangan (NMPC) to D. B. Vassallo (NRC) dated May 7, 1985.
5.
NSAC-21. "Fundamental Safety Parameter Set for Boiling Water Reactors".
6.
Regulatory Guide 1.97 (Revision 2), "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980.
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