ML20245C676
| ML20245C676 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 04/21/1989 |
| From: | Chrzanowski R COMMONWEALTH EDISON CO. |
| To: | Murley T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20245C675 | List: |
| References | |
| GL-83-28, NUDOCS 8904270332 | |
| Download: ML20245C676 (5) | |
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"\\ Commonwe:lth Edison II 72 West Adams Street, Chicago, llknois
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" / adrEs Reply to: P6sTDftsiB6i767-i
' s._ / Chicago, Ilhnors 60690 0767 April 21, 1989 Dr. Thomas E. Murley, Director Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Byron Units 1 and 2 Braidwood Units 1 and 2 Generic Letter 83-28 Items 2.2 (Part I), 4.2.3 and 4.2.4 HRC Docket _Hos. 50-454/455 and 50-456/457 References (a) November 5, 1983, letter from P.L.
Barnes to H.R. Denton (b) February 29, 1984, letter from P.L. Barnes to H.R. Denton (c) Mary 5, 1988, letter from L.N. Olshan to H.E. Bliss (d) September 9, 1987, letter from D.R. Muller to L.D. Butterfield
Dear Dr. Murley:
Reference (a) and (b) provided Commonwealth Edison's response to Generic Letter 83-28.
Reference (c) requested additional information from Byron and Braidwood on Item 2.2 (Part I).
Attachment A to this letter contains the response to the questions regarding Item 2.2.
Reference (d) i requested additional information from Byron and Braidwood on Items 4.2.3 and l
4.2.4.
Attachment B to this letter contains the response to the questions regarding Items 4.2.3 and 4.2.4.
Please address any further gitestions concerning this subject to this office.
Very truly yours,
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R.A. Chr anowski 8904270332 890421 PDR ADOCK 0500045f P
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Resident Inspector-Byron l
Resident Inspector-Braidwood L.N. Olshan-NRR l
S.P. Sands-NRR l
D.R. Muller-NRR
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Region III Office 1
lL._________.-_._-_.-_-_-__.
4 ATTACHMENT A LThe following information is being submitted in response to the request for additional information for Items 2.2.1.3 and 2.2.1.4.
NRC Requgst for Item 2.2.1.3 (Use of Information Handling System)
The licensee should identify and describe the procedure by which supervisory and other personnel are required to use the information-handling system to determine the classification of both the activity and the components involved in the activity.
In addition the
'llcensee should confirm that classification of an activity as safety-related automatically causes the use of safety-related procedures, cautions and constraints.
EREPMS21 Supervisory and other personnel are required to use the information handling system to determine the classification of activities and components at Byron and Braidwood Stations.
Specifically, the Operating Engineer is responsible for classifying maintenance activities as safety-related by Administrative Procedure AP 1600-1,
" Initiating and Processing a Nuclear Work Request (NWR)".
In addition to the Operating Engineer's classification of an NWR, Work Analysts, Quality Control (QC), Quality Assurance (QA) and Technical Staff In Service Inspection (ISI) personnel are required by AP 1600-1 to verify the safety-related classification.
When an NWR is classified as safety-related by an Operating Engineer, AP 1600-1 ensures that safety-related procedures, cautions and constraints are utilized.
Safety-related NWR's require additional levels of approval and review, and require the use of the information handling system in the development of detailed work instructions.
Lead Workman are required to document troubleshooting and disassembly of safety-related equipment on approved administrative procedure forms.
The Quality Assurance Manual Quality Procedure (QP) 2-53 " Quality Assurance Program for Operations Classification of Structures, Systems and Components" provides the direction for supervisory and j
other personnel on the use of the information handling system (Engineering Lists) to determine the classification of components and thus the activities associated with the installation of these components. This OP specifies that PWR Engineering is responsible for providing complete listings of components, piping, instrument and j
wiring diagrams to identify safety-related itemr for use in procurement, maintenance, repair, modifications and for ordering parts and meterials.
Examples of these Engineering Lists include the equipment 0-List, Valve List, Instrument Index List.- Mechanical Equipment List and the As-Built P3 ping and Instrument Diagrams.
These lists have been compiled into the Safety Related Component List which is used as the basis for determination. This QP states that these lists will be used in classifying work requests specified in QP 3-51 and QP 3-52 used for performing plant modifications and plant maintenance.
Per these OP's the responsibility for classifying work requests as to whether work is maintenance or a modification, and whether ASME, Regulatory or safety-related and whether class 1E equipment is involved lies with the Operating Engineer.
l ATTACHMENT A (continued)
NRC Requggt for Item 2.2.1.4 (Manage. ment _ Controls )
The licensee should describe their program of management controls-that will be used to assure their management that the equipment classification information handling system was properly prepared and validated, is properly maintained, and is being used as intended.
j Responset l
PWR Engineering has responsibility for the preparation and maintenance of the Safety-Related Component List (SRCL) for Byron and Draidwood Stations._ This task is governed by the Ceco QA manual and Quality Procedures, specifically, O.P.
3-3, Classification of Systems, Components, Parts and Material; Q.P. 3-51, Design Control for Operations - Plant modification; Q.P. 4-51, Procurement Document Control for Operations - Processing Purchase Documents.
- Also, Engineering Procedure 0 12.4, Control and Maintenance of the Safety-Related Component List (SRCL) for Byron and Braidwood Stations, provides detailed direction to the engineer for all hardware changes which involve the addition, deletion, non-identical replacement or relocation of safety-related components, or non-safety-related components associated with safety-related systems and results in the preparation and maintenance of the Safety-Related Component List.
The SRCL is a controlled design document identifying safety-related components, non-safety-related components in safety-related systems and components in Byron and Braidwood Stations.
For Byron and Braidwood Stations, the PWR Engineering Manager has delegated responsibility to Sargent & Lundy for the control, issuance, and j
maintenance of the SRCL to be implemented in accordance with S&L Project Instruction PI-88-68.
The PWR Engineering Manager assures that necessary changes to SRCL are forwarded to S&L when modifications to safety-related components are designed by A-Es other than S&L.
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ATTAQRiENT B This response provides Commonwealth Edison's justification for not establishing a predetermined life or change out schedule for the trip and bypass (DS-series) breakers located in the reactor trip switchgear enclosure.
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The Byron and Braidwood electrical maintenance procedures, for the switchgear 1
assembly, are updated in a timely manner in accordance with various industry requirements. The Station's operating, electrical maintenance, and tech staff personnel conduct both bi-monthly and refueling outage interval surveillance. Data from breaker counter readings, system initiated trip time responses, Undervoltage Trip Attachment (UVTA) and Shunt Trip Attachment (STA) initiated trip time responses, and applied UVTA threshold voltages to trip are trended by the Technical Staffs upon complett a of the surveillance.
If a trended parameter becomes off-normal, then a nuclear work request is issued and the breaker is re-examined.
The WCAP 10835 report, for life cycle and compatibility testing of trip attachments, utilized six new DS-416 RTS breakers for testing the undervoltage and shunt trip attachments. These breakers acewnulated 44,107 operations, at no ampere load, with an average of 7,350 operations per breaker and never failed to trip.
Breaker maintenance for the test was performed after 600 cycle intervals. The same maintenance is performed at the stations each refueling outage.
Since Byron and Braidwood experience an average of 14.5 operations per month for the reactor trip breaker, and a much smaller rate per month for the bypass breaker, the same maintenance would be performed within an eighteen month, 261 cycle interval.
This accelerated maintenance practice is performed at about one half of the vendor recommended period of 500 cycles.
Also, Byron and Braidwood Station's commitment requires replacing the trip attachments at 1250 cycle intervals.
The 14.5 operations per month is the average monthly trip break--
operations derived from breaker counter readings taken from both stations.
Initial readings were taken in October, 1987.
Follow-up readings varied from 14 to 17 months and contained an average monthly trip breaker range of 9 to 21.8 operations per month.
A 40 year plant life projection of 7,614 breaker operations was developed by using the individual breaker monthly rates to find the number of operations at the time of each units service date and then using the overall average of the breaker monthly rates to determine a common projection for 480 months.
ANSI C37.16 - 1988 is a standard for all low voltage, 600 VAC and lower, power circuit breakers and AC power circuit protectors that provides information on preferred ratings, related requirements, and application requirements. The endurance parameters from this standard require 3200 breaker operations at no ampere loading and 800 breaker operations at rated 1600 ampere load for a total of 4,000 operations. The projection from our present surveillance / maintenance activities will result in a doubling of the ANSI endurance level.
ATTAClaiEtiT B (continuedl This standard addresses low voltage AC circuit breakers that are arranged in straight bus with circuits for large feed type breakers and smaller load type breakers. The reactor trip switchgear breakers are also low voltage AC switchgear breakers but are arranged in ladder fashion of two trains with trip and bypass breakers for each train.
By design, these breakers are never subjected to interrupting electrical faults and provide the final control point for dropping rods. The upstream motor-generator set output breaker will interrupt electrical power system type faults in the feeds associated with the reactor trip switchgear. This hybrid usage should be less stressful to the breaker internals and ongoing trending would identify problems as they begin.
Based on the above program description, it is Commonwealth Edison's position that a high margin of reliability can be maintained without a breaket replacement program, i
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