ML20078Q967

Forwards Response to Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events. Response Describes Status of Conformance W/Positions of Generic Ltr & Schedules for Needed Improvements

ML20078Q967
Person / Time
Site:	Summer
Issue date:	11/04/1983
From:	Dixon O SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
GL-83-28, NUDOCS 8311140241
Download: ML20078Q967 (10)

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SOUTH CAROLINA ELECTRIC & GAS COMPANY Post oarica 7e4 Cotuus A. SOUTH CAROUNA 29218 O W. DIXON, JR.

uve$.'d"o',*.'.57[o , November 4, 1983 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Generic Letter 83-28

Dear Mr. Denton:

This letter is provided in response to Generic Letter 83-28,

" Required Actions Based on Generic Implications of Salem ATWS Events." The attachment describes the status of current conformance with the positions stated in the Generic Letter and include plans and schedules for needed improvements as appropriate. For convenience, the items are addressed in order of their appearance in the Generic Letter.

Pursuant to 10CFR50.54(f), the undersigned attests that the information provided herein is true and correct.

If you should have any questions, please call us at your convenience.

l Very truly yours, (h &

8311140241 831104 O. W. D on, r.

PDR ADOCK 05000395 P PDR AMP:OWD/fjc Attachment cc: V. C. Summer A. R. Koon T. C. Nichols, Jr./O. W. Dixon, Jr. C. L. Ligon (NSRC)

E. H. Crews, Jr. G. J. Braddick E. C. Roberts D. J. Richards H. N. Cyrus NRC Resident Inspector J. P. O'Reilly J. B. Knotts, J r~.

General Managers g NPCF W. A. Williams, Jr. lgh\ File (Lic./Eng.)

O. S. Bradham N. E. Clark

{v sI C. A. Price 6

Generic Letter 83-28 November 4, 1983 Page #1 ATTACHMENT POST TRIP REVIEW l.1 As stated in the letter from O. W. Dixon, Jr. to H. R. Denton, dated July 15, 1983, South Carolina Electric and Gas Company (SCE&G) has established a program for performing Post Reactor Trip Reviews.

This program is addressed and implemented by Station Administrative Procedure (SAP)-132, "Off-Normal Occurrence Evaluation, Reporting and Resolution,"

.and Station Emergency Operating Procedure (EOP)-5,

" Recovery From Reactor Trip." All items required by this section of the Generic Letter are included in these procedures with the exception of item 1.l(3).

This item, dealing with the qualifications and training of responsible personnel, is addressed in the Final Safety Analysis Report, Chapter 13, and is in accordance with Technical Specifications 6.3., " Unit Staff Qualifications," and 6.4, " Training" requirements.

1.2.l(1) The capability for assessing the Sequence of Events (SOE) surrounding a reactor trip is provided by the Westinghouse Process Computer P2500.

1.2.l(2) A sequential events recording function is pro-vided to monihor the exact sequence of operations to permit analysis of an event. When not in an event sequence,. normal SOE inputs (non-trigger) are processed as discrete inputs with at least a one second scan frequency. An event sequence is initiated when any of the inputs specified in the SOE trigger field of the database change state.

The system then monitors approximately 90 key parameters with a recording resolution of 3 milliseconds.

l.2.l(3) The resolution in time on the SOE Record print-out is to the nearest cycle. A chronological record of events is provided with an accuracy of approximately 3 milliseconds. If the P2500 detects two inputs which have changed state within the 3 millisecond window, asterisks will be displayed to indicate that the P2500 could not determine which of the events occurred first.

Generic Letter-83-28 1 Attachment November 4, 1983 LPage #2 1.2.1(4)' The Sequence of Events Record is preceded by two blank lines to visually offset its printout from.the rest of the Alarm Typewriter printout. The first line of the printout.provides the current time of day, identifies the-subsequent printout as the SOE Record, and gives the. time of the change of state which caused the program to be triggered (ie., the First Event). First Event Time is displayed to the nearest second. In.the' ensuing printout of digital

, points, time'is printed as cycles referenced to the First Event. Time. Any of the SOE digitals which change state in the one-minute period following the First Event will be included in the'SOE Record.

Additional 3600 cycle SOE Records are generated when trigger inputs change state. The, printout is terminated by'a printout line which'provides the current time of day and identifies the line as "END OF SOE RECORD."

1.2.l(5) The P2500 Process Computer stores the input information described above until it is printed.

The printouts produced provide a historical record.

1.2.1(6)- The normal power source for the P2500 Process Computer is a DC.to AC inverter 5905 which is fed by

.a battery backed Non-Class lE DC circuit. The alternate power source is a Non-Class lE 120VAC bus.

1.2.2(1) The Post Trip Review capabJlity is a feature of the P2500 Process Computer which provides a log of the time-dependent behavior of significant selected analog inputs before and'after a reactor trip and/or safety injection. The Post Trip Review printout, in conjunction with the SOE Record printouts, provides diagnostic information necessary to determine the cause of a. reactor trip and to determine whether plant performance following a trip was as expected.

~1.2.2(2)- The P2500~ Process Computer Program continuously stores.a two minute accumulation of selected plant analog inputs which are scanned at ten second intervals. -The inputs include those necessary for (1) event diagnosis, (2) performing immediate preplanned operator action and (3) taking the plant to a-safe shutdown condition. It also stores a ten i

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Generic Letter 83-28 Attachment November 4, 1983 Page #3 1.2.2(2) (Continued) second accumulation of a few plant analog inputs which are scanned at 2.5 second intervals. When a reactor trip or safety injection occurs, the program stores an additional three minutes accumulation of the ten-second scan analog input data and an additional ten seconds accumulation of the 2.5 second scan analog input data. At the end of the three-minute post trip accumulation period, the program prints the data on the Line Printer or on the Trend Typewriter if the Line Printer is ou t-of-s ervice.

1.2.2(3) The duration of the stored time history includes two minutes before and three minutes after the post trip data acquisition sequence is triggered.

1.2.2(4) The first line of the first page of the printout provides the current time of day that the Post Trip Review printout started, and provides the date and time that the trip or safety injection occurred (ie., Trip Time). Trip Time is printed to the near-est tenth of a second. In the following printouts of analog point data, time is printed as minutes, seconds, and tenths of seconds. The " hours" are only printed once in the Trip Time and not printed on each page of the printout. The final line of the final page of the Post Trip Review printout provides the time at which the printout is completed.

1.2.2(5) The P2500 Process Computer will retain stored analog data as described in Section 1.2.2(3). The printouts produced provide a historical record.

1.2.2(6) The power sources are the same as described in Section 1.2.l(6).

1.2.3 Other data and information which are available include applicable Recorder Traces (ie., Pressurizer Pressure, Pressurizer Level, RCS Temperature, etc.)

as well as information stored in the Technical Support Center Computer.

1.2.4 There are no planned changes to existing systems.

7 Generic Letter 83-28 ATTACHMENT November 4, 1983 Page #4 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE 2.1 The components of the Reactor Trip system are described in FSAR Section 7.2. Ts list has been reviewed and the component classifications have been verified to be properly identified in the documentation system utilized to control safety related components as described in sections 2.2.l(1) and (2) of this letter. The majority of the Reactor Trip System components were provided by Westinghouse as a part of the NSSS system. Baseline technical information documents for these components were provided with the equipment by Westinghouse. Supplemental and revised information has been, and continues to be, supplied via Westinghouse Technical Bulletins.

Distribution of Westinghouse Technical Bulletins now includes a return receipt. The return receipts are pre-addressed to a central point in Westinghouse for recording all Technical Bulletins transmitted and their status. Technical Bulletins for which receipts are not acknowledged within a reasonable time are retransmitted. A list of current Technical Bulletins has been provided and will.be updated periodically but not less frequently than once per year. These updates are in the form of a Technical Bulletins.

All technical information received by SCE&G from vendors which has a direct impact on plant design, maintenance, and safe operation is distributed to appropriate organizations for action. Additionally, this vendor technical correspon-dence is tracked for proper distribution and disposition utilizing a computerized tracking system. The Westinghouse Technical Bulletin Program along with the SCE&G procedure for handling vendor technical correspondence provides a complete vendor interface program for the Westinghouse supplied Reactor Trip System components.

The vendor interface program for those components in the Reactor Trip Sy' tem not supplied by Westinghouse will be conducted in accordance with the program to be developed for other safety related components.

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' Generic Letter 83-28 ATTACHMENT

November 4, 1833 ,

'Page: 95, i

- 2.2.l(1) . Equipment Classification for all systems has been accomplished as. described in FSAR Section 3.2. Lists ofJaystems and equipment classified as safety-related

- .are contained in FSAR Sections-3.2 and 3.11. The SCEEG Operational QA Program requires that any

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materials integral to or in contact with these systems,' components, or structures are. treated

. safety related unless declassfied to non nuclear safety by' Engineering under the design control program.

2.2.l(2)- In addition to the equipment lists in FSAR Chapter 3, engineering design documents such as drawings, specifications and bills of material which have been reviewed and approved under the plant design control

program'are issued with safety related designations.

.At the present time, the safety classification field of the Computerized History and Maintenance Planning System (CHAMPS) . listing is being validated. When the validation has been completed, which is scheduled for January 1984, this computerized listing will be utilized _to identify the safety classification of components. The present system will remain as a back-up'to the computer system.

2.2.l(3) When activities defined in the introduction to 10CFR50, Appendix B, are to be performed, both the organization which will perform an activity and the Quality Services organization verify. proper classification of work procedures. This is performed by checking approved. design documents thus assuring procedures appropriate to the safety classification are used. At this time, work or procurement package documentation is also reviewed for compliance to the design documents, FSAR, approved drawings, etc. If the safety classification of the affected components can not be clearly determined, a disposition ccncerning the subject activity is provided by-Engineering.

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Generic-Letter 83-28 ATTACHMENT November 4, 1983 Page #6 2.2.l(4) The-management controls utilized to verify that the procedures utilized in performing activities associated with safety related components are as specified in the FSAR Chapter 17 and Section 6 of the Technical Specificatons and described in the Operational QA Plan and associated procedures. These controls consist primarily of QA audits and surveillances.

2.2.l(5) When replacement equipment or parts are ordered for safety related equipment, they are procured via safety related procurement documents. Engineering specifies the technical requirements, including design verification and qualification testing, included on these documents. The entire purchase requisition package is reviewed by QA prior to a purchase order being placed. If parts or equipment which have been procured via non safety related purchasing procedures are needed for use in a safety related application, they are dedicated in accordance with criteria established by Engineering and reviewed by QA before they can be installed.

These processes are described in appropriate i procedures. l 2.2.1(6) The equipment classification program is as described in FSAR Section 3.2 2.2.2. As stated in the letter from O. W. Dixon, Jr. to H. R. Denton, dated September 2,1983, SCE&G intends to develop a vendor interface program in accordance with the recommendations of the Nuclear Utility Task Action Committee, coordinated by INPO, which has been established to address tais item.

e Generic Letter 83-28 ATTACHMENT November 4, 1983 Page #7 POST MAINTENANCE TESTING '

3.l(l) Applicable procedures have been verified to require that post-maintenance operability testing of Reactor Trip System components is conducted in accordance with approved test procedures and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

.3.l(2) To ensure that any appropriate test guidance is included in the test and maintenance procedures, a check of vendor and engineering recommendations concerning Westinghouse supplied Reactor Trip System components has been performed. Recommendations have been properly incorporated into appropriate test and maintenance procedures with the exception of the recommendation outlined NSD-TB-77-ll, dated July 21, 1977. The testing described in this technical bulletin is being discussed with Westinghouse to determine applicability.

, 3.l(3) No changes to any post maintenance testing requirements are required at this time.

3.2(1) Applicable procedures have been reviewed and post maintenance operability testing is performed as appropriate, subsequent to maintenance activities, on safety related equipment identified in the Technical Specifications.

3.2(2) A check for incorporation of vendor and engineering recommendations concerning test guidance for safety related components other than the Westinghouse supplied Reactor Trip System components will be performed comensurate with the schedule developed in response to item 2.2.2.

3.2(3) The response to this item is identical to the response to 3.l(3).

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Generic Letter 83-28 ATTACHMENT' November 4, 1983 Page #8 ,

REACTOR TRIP SYSTEM RELIABILITY 4.1 A review of vendor recommended reactor trip breaker modifications has been performed and each modification, with the exception of the automatic actuation of the breaker shunt trip attachment, has been implemented.

. 4.2(1)&(2)The preventive maintenance and surveillance

' program for' the reactor trip breakers is as described in the O. W. Dixon, Jr. to H. R. Denton letter, dated June 8, 1983.- This program allows for the trending of parameters affecting operation by recording critical measurement during surveillance testing.

4.2(3)&(4)The Life Cycle Testing Program for the reactor trip breakers was addressed in'the O. W. Dixon to H. R. Denton' letter dated June 30, 1983. Life cycle testing of the shunt trip attachment and the undervoltage . trip attachment of the~ reactor trip switchgear is being conducted by Westinghouse-for the Westinghouse Owners Group. This program is directed toward establishing the service life of these devices, and substantiating periodic test requirements with proper maintenance. The results of this program will lun factored into maintenance, replacement and qualification programs. The test program is scheduled for completion in the second quarter of~1984.

i 4.3 The automatic reactor trip system actuation cap-I ability of the breaker shunt trip attachments will be installed as described in letters from i O. W. Dixon, Jr. to H. R. Denton dated August 31, 1983.and October 20, 1983. When approved by the L NRC, this modification will be installed during the inext outage of sufficient duration.

4.4 Not applicable I

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Generic Letter 83-28 ATTACHMENT November 4, 1983 Page 99 4.5(1) As described in'the O. W. Dixon, Jr. to H. R. Denton letter dated May 12, 1983, the present design does not allow--independent testing of the undervoltage and shunt ~ trip devices. A design change to allow this testing has been submitted in O. W.:Dixon, Jr..to 1H. R. -Denton letters dated August 31 and October 20, 1983..When approved, this design change will be in-stalled during the next outage of sufficient duration.

4.5(2) Nob applicable.

4.5(3) In January 1983, the Westinghouse- Owners Group sub-mitted WCAP-10271 to the NRC for review. WCAP-10271,

" Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumenta-tion. System," documents an evaluation of the impact of current and extended surveillance intervals on RPS unavailability.

The WCAP considers common-node failure, operator error, reduced redundancy'during testing'and equipment bypass.

WCAP-10271 also considers correlative effects on plant operation and. safety including the manpower expenditure associated with surveillance, the number of inadvertent trips which occur during testing,-and the distraction from plant monitoring on the part of the control. room l operator and shift supervisor associated with testing.

Supplement 1 to WCAP-10271 submitted to the NRC in September 1983, is an extension of the evaluation and

! provides a discussion of component wearout caused by

[ testing. The NRC review of WCAP-10271 resulted in a p request for additional information the NRC felt i necessary to complete the ' review. Information was l submitted to the NRC in response to the request and included an overall' evaluation of the impact on plant safety of RPS surveillance, a discussion of the uncer-y tainty of failure rates and common mode failure, and more detail concerning the impact of surveillance intervals on RPS unavailability. WCAP-10271, Supple-

, ment 1 to WCAP-10271, and the information provided to the NRC in' defense of WCAP-10271 provides in a compre-L hensive form the information requested by this item.

i- The conclusion of WCAP-10271 and Supplement 1 is that although RPS unavailability is increased, less frequent testing of RPS components is warranted and will result

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in an improvement in overall plant safety and

. equipment reliability.

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