ML20023C043

Affidavit of Dr Woodlan Providing Responses to ASLB Inquiries Re Board Notifications 83-26 & 83-38 Concerning Failure of Certain Reactor Trip Breakers to Perform as Designed.Prof Qualifications Encl

ML20023C043
Person / Time
Site:	Comanche Peak
Issue date:	05/05/1983
From:	Woodlan D TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20023C035	List: ML20023C034 ML20023C037 ML20023C039 ML20023C040 ML20023C041 ML20023C042 ML20023C043
References
NUDOCS 8305090530
Download: ML20023C043 (12)

Text

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

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TEXAS UTILITIES GENERATING ) Docket Nos. 50-445 COMPANY,et- al. ) 50-446

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(Comanche Peak Steam Electric ) (Application for Station, Units 1 and 2) ) Operating Licenses)

AFFIDAVIT OF DONALD R. WOODLAN REGARDING REACTOR TRIP BREAKERS AT COMANCHE PEAK I, Donald R. Woodlan, being first duly sworn, do depose and state as follows: I am employed by Texas Utilities Services, Inc. as a Senior Licensing Engineer. In this position I am responsible for the licensing matters relating to electrical and instrumentation and control systems and equipment for Comanche Peak. As such, I am familiar with Applicants' commitments regarding reactor trip breakers. A statement of my professional qualifications is attached hereto.

This affidavit provides information in response to the Board's inquiries regarding Board Notifications 83-26 and 83-38 concerning failure of certain reactor trip breakers to perform as designed. In addition, this affidavit addresses the Board's 8305090530 830505 PDR ADDCK 05000445 T PDR

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questions regarding the status of compliance with certain Staff positions set forth in NUREG-0460 and IEEE-279. Specifically, I address the following questions:

(1) Does Comanche Peak utilize either type of reactor trip breaker referenced in Board Notifications 83-26 and 83-38?

(2) If not, what type of reactor trip breakers are used at Comanche Peak?

(3) What maintenance or other procedures have been implemented at Comanche Peak to provide reasonable assurance that an ATWS event will not occur?

(4) What is the status of Applicants' position regarding diversity of scram breakers in Westinghouse facilities?, and (5) What is the status of Applicants' satis-l faction of IEEE-279 regarding protection systems for nuclear power generating stations?

I. Board Notifications 83-26 and 83-38:

Failure of Reactor Trip Breakers The Board has inquired whether the reactor trip breakers employed at Comanche Peak are those referenced in either Board Notification ("BN") 83-26, " Failure of Reactor Trip Breakers to Open on Trip Signal" (March 3, 1983) or BN-83-38, " Failure of

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GE AK-2 Reactor Trip Breakers" (March 22, 1983). 1 As previously addressed by Applicants 2, Comanche Peak does not employ the Westinghouse DB-50 reactor trip breakers discussed in BN-83-26. Neither does Comanche Peak employ the GE AK-2 reactor trip breakers which were the subject of BN-83-38.

Nevertheless, because each of the failures described in BN 26 and BN-83-38 may have resulted from inadequate maintenance of breakers, Applicants have reexamined their maintenance procedures for these components, as discussed below.

II. Reactor Trip Breakers Employed at Comanche Peak Comanche Peak utilizes Westinghouse DS-416 reactor trip breakers. Although no maintenance-related failures have been observed with these breakers, Westinghouse has recently documented the misoperation of a DS-416 breaker during pre-planned testing of the undervoltage trip function of the breaker at an operating plant. Westinghouse notified Applicants of this failure by letter dated April 20, 1983 (Attachment B). As described in that letter, the cause of the failure was evidently manufacturing variations and/or design discrepancies between certain components of the undervoltage 1 April 7, 1983 Conference Call, Tr. 16-17.

2 See " Applicants' Assessment of Relevance and Significance of Board Notifications" (March 18, 1983) at 15-16.

trip attachment. Westinghouse has committed to replace all such undervoltage trip attachments with attachments modified to correct the identified manufacturing variations. See Attachment B. These new breakers will maintain the high reliability of operation provided by the Westinghouse reactor 4 protection systems.

III. Maintenance Procedures Following the reactor trip breaker malfunction at the Salem facility, Westinghouse issued a technical bulletin applicable to testing procedures for the DS-416 breakers. In response to that bulletin, Applicants reviewed their test procedures for those breakers. Because Applicants had previously identified the need to verify that both manual and undervoltage trip mechanisms were operable in tests (as recommended in the Westinghouse technical bulletin), no changes to those procedures were necessary.

With respect to maintenance procedures for the DS-416 breakers, Applicants have reviewed their procedures in response

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j to I&E Bulletin 83-01, issued in response to the Salem breaker malfunction. Applicants' procedures already provide for annual maintenance of the DS-416 breakers. The maintenance schedule is controlled by the plant maintenance program computer, and a notice of required maintenance will be issued. The maintenance 1

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procedures call for cleaning, inspection and lubrication of the breakers. Accordingly, Applicants have determined that their maintenance program is adequate.

IV. Diversity of Reactor Trip Breakers ATWS has been an issue in the licensing of nuclear power reactors since 1969, and I need not recount that background here. 3 In addition, the NRC has embarked on a rulemaking to

) establish requirements for resolving ATWS. The history of this rulemaking is described by the NRC Staff in the Affidavit of David W. Pyatt (at 5-8), attached to the Staff's May 7, 1982, Answer to Board Question 3. A final rule is expected in June 1983. Upon adoption of a final rule, Applicants will of course comply with the provisions applicable to Comanche Peak.

In response to the specific question of diversity of scram i breakers in Westinghouse facilities which has been raised by the Board 4, Applicants note that the Comanche Peak FSAR l

l provides an extensive discussion of the design of the reactor trip system. See FSAR {7.2. The advantages and disadvantages of adding additional diversity to the reactor trip breakers was L

l l 3 The NRC Staff summarized the history of the ATWS issue in its May 7, 1982, " Answer to Board Question 3, Regarding j

the Status of Safety Issue TAP A-9" at 3-8.

4 April 7, 1983 Conference Call, Tr. at 17-18.

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discussed in NUREG-0460. 5 Westinghouse has demonstrated that adequate diversity exists in the Westinghouse designs and additional diversity in the reactor trip breakers is not required. In particular, with respect to common mode failures for reactor protection systems, measures such as functional diversity, physical separation and testing, as well as administrative controls during design, production, installation and operation have been taken. FSAR $7.2.2.2.3, p. 7.2-30.

V. Satisfaction of IEEE-279 Criteria As noted above, IEEE-279 6 provides criteria for reactor protection systems. The criteria established there contain both requirements and recommendations for such systems. The reactor protection systems for the Comanche Peak facility satisfy all requirements set forth in IEEE-279. Indeed, the licensing basis for the Comanche Peak construction permit included a commitment to IEEE-279 requirements. See PSAR 7.2 as presented in RESAR-3, $7.2.2.3. 7 Thus, the design of the 5 See NUREG-0460, Volume 4, Section 2.3.4.1.

6 IEEE-279, "IEEE Standard: Criterion for Protection Systems for Nuclear Power Generating Stations" (June 3, 1971).

7 RESAR-3 is the Westinghouse Reference Safety Analysis Report for the Westinghouse generic pressurized water reactor design. RESAR-3 was referenced in the Comanche Peak PSAR.

reactor protection systesa for Comanche Peak, which began construction in 1974, has always natisfied IEEE-279 requirements.

VI. Conclusion In conclusion, the reactor protection system at Comanche Peak, including the reactor trip breakers, has been designed to satisfy all applicable NRC requirements. In addition, the l

i l reactor protection system for Comanche Peak meets IEEE-279 requirements, and Westinghouse has demonstrated that

! additional diversity in the reactor trip breakers need not I

be implemented for Westinghouse facilities.

I Donald R. Woodlan County of Dallas )

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State of Texas )

Subscribed and sworn to before me this 5th day of May, 1983.

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/ wk Notary Public A d' Y GLENCA Suit) i. M:..: .

b c.nd for D3!;as caer.r g,,

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This is a telecopy facsimile. The original will be trans-mitted under separate cover.

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ATTACHMENT A 2 DONALD R. WOODLAN STATEMENT OF EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS POSITION: Senior Licensing Engineer FORMAL EDUCATION: 1964-1968, B.S. Electrical Engineering and Mathematics, United States Naval Academy 196 8-196 9, M.S. Electrical Engineering, Michigan State University PROFESSIONAL REGISTRATION: Ohio - Professional Engineer EXPERIENCE:

1979 - Present Texas Utilities Services Inc. as Senior Licensing Engineer: Comanche Peak Steam Electric Station.

Responsible for the I&C (instrument and control) and Electrical aspects of licensing, and for electrical equipment environmental qualification.

1975 - 1979 Cleveland Electric Illuminating Co.

as Operations Engineer, Perry Nuclear Power Plant responsible for spare parts program, technical specifica-tions, system operating description, and control panel layouts.

1968 - 1975 U.S. Navy as Division Officer, Department Head, Engineering Watch Officer, Ship Duty Officer: Nuclear powered submarines (USS Jack and USS Sam Rayburn). Responsible for the operation and maintenance of nuclear l

reactor and ships power plant; for the safe operation of the ship; and for the operation and maintenance of l the weapons systems.

PROFESSIONAL SOCIETIES: Institute of Electrical and Electronic j

Engineers - Member American Nuclear Society - Member i

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, f g ATTACHMENT B (G

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Nxtear Operations 0msion Westinghouse. Water Reactor

( Electric Corporation Divisions ex 355 Pmsburgh Pennsytvania 15230 April 21,1983

. WPT - 6249 Mr. J. B. George Vice President / Project General Manager Texas Utilities Services , Inc.

P.O. Box 1002 Glen Rose, Texas 76043 TEXAS UTILITIES GENERATING COMPANY COMANCHE PEAK STEAM ELECTRIC STATION DS-416 REACTOR TRIP SWITCHGEAR

Dear Mr. George:

Confirming the telecon on April 20, 1983, between J. Marshall (TV) &

M. Toracso (W), Westinghouse advised the Nuclear Regulatory Commission of the potentiaT for misoperation of DS-416 reactor trip switchgear undervoltage attachments. This was reported to the NRC under 10CFR21 for operating plants, and under 10CFR50.55(e) for plants under construction.

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Westinghouse was recently informed by a utility that one Model DS-416 main reactor trip breaker at each of two plants did not trip during pre-planned testing of the undervoltage (UV) trip function.

! Based upon a review at the plant site and at Westinghouse, the following

, items have bee'n identified as the factors potentially involved in the

! reported occurrences :

a. Manufacturing variations permitted interference between the moving core and the roller bracket on one of the UV devices. A further factor may have been lack of the side-to-side clearance of the roller bracket.

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b. Lack of minimum gap between the UV trip reset lever and the breaker trip bar pin appears to be related to the malfunction of the second UV device.

( The Westinghouse evaluation has concluded that deviations from the recommended clearances could increase the potential for misoperation of the attachment, thereby creating a condition wherein the reactor trip switchgear might not open on automatic demand from the reactor protection system.

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Page 2 WPT - 6249

( Subsequently, Westinghouse advised the operating plants on April 15,1983, of an additional misoperation of another DS-416 undervoltage attachment. Investigation of this event' revealed a missing retaining ring on one of the two undervoltage attachment pivot shafts (shown in Attachment 1). This allowed the pivot shaft to move laterally such that one end came out of its guide hole in the frame of the undervoltage attachment, and did not permit the attachment to operate on demand. A missing retaining ring was also identified at another plant.

No misoperation of that attachment had been reported.

The Westinghouse evaluation of the retaining ring issue revealed a discrepancy in design. The groove in the shaft receiving the retaining ring was not increased in width to be consistent with an earlier (1972) retaining ring design change. The new retaining ring is wider than the original design and does not seat properly in the existing grooves. This discrepancy increases the potential for misoperation of the DS-416 undervoltage attachment, thereby creating a condition wherein the reactor trip switchgear might not open on automatic demand from the reactor protection system.

Corrective Actions (1) Westinghouse is committing to its utilities to replace the undervoltage attachments on DS-416 reactor trip switchgear supplied by Westinghouse for its Nuclear Steam Supply Systems. (2) The new attachments have modified (widened) grooves to accomodate the new retaining rings. (3) Manufacturing drawings have been revised and quality control procedures modified to assure that

( critical design dimensions are maintained during manufacture. (4) Furthermore, Westinghouse is developing and will implement a procedure for installation of the new attachments on DS-416 reactor trip switchgear in the plant. This

[ field installation procedure will provide for proper alignment and interface l

of the attachment with the breaker trip shaft.

l To justify continued plant operation until replacement undervoltage attachments can be fabricated, shipped, and properly installed, Westinghouse recommends the following be performed:

(a) Assure that control room operators are alerted to the potential for misoperation of the reactor trip switchgear due to possible malfunction of the undervoltage attachments.

j (b) Re-emphasize to the control room operators the indications available to

! detect failure of the rods to insert into the core following a demand l

signal from the automatic protection system.

l (c) Re-emphasize to the control room operators the manual reactor trip options available in the existing emergency operating procedures.

These procedures contain diverse methods of tripping the reactor which do not rely on the undervoltage t' rip attachments.

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. WPT - 6249

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(d) Visually verify after operation of the reactor trip switchgear that retaining rings are located on the grooved ends of the two under-voltage pivot shafts identified in Attachment 1. This will provide increased confidence that the shaf t will not disengage, and that the attachment will be operable for the next trip demand. After installation of the replacement DS-416 undervoltage attachments , the requirement to visually verify the presence of retaining rings after operation of the reactor trip switchgear can be eliminated.

If you require additional information on this, pleas ~e contact me.

Very truly yours ,

WESTINGHOUSE ELECTRIC CORPORATION 77 h Y m S A. T. Parker, Manager 14RD Comanche Peak Project ATP/mkv cc: J. B. George IL I. J. T. Merritt il R. D. Calder IL H. C. Schmidt 1L R. E. Ballard IL C., B . Hartong il J. C. Kuykendall lL G. C. Creamer IL ARMS IL l R. A. Jones 1L.

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, bec: A. T. Parker il F. M. McDonough lL.

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, NSD PTant Date File, 701 Bldg. lE

R. L. Moller TBX/TCX Site Mgr. 2L R. W. Skoff ll, l A Bobbie Kendig ITTC Tech Library ll.

M. A. Torcaso 1L E. R. Strussian ll 4

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