ML20211P636

Application for Amend to License NPF-11,authorizing Operation W/Crd 10-47 Withdrawn for Remainder of Cycle 2.No Significant Hazards Exist.Fee Encl

ML20211P636
Person / Time
Site:	LaSalle
Issue date:	12/15/1986
From:	Allen C COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20211P641	List: ML20211P636 ML20211P713 ML20211P747
References
2519K, NUDOCS 8612190112
Download: ML20211P636 (5)

Text

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f Commonwealth Edison S t , One First National Plaza. Chicago, Illinois

\ ~ Address Reply to: Post Office Box 767 N/ Chicago, Illinois 60690 0767 December 15, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

LaSalle County Station Unit 1 Proposed Amendment to Technical Specification for Facility Operating License NPF To Allow Operation with CRD 10-47 Withdrawn One Cycle NRC Docket Nos. 50-373

Dear Mr. Denton:

Pursuant to 10 CFR 50.90, Commonwealth Edison proposes to amend Appendix A, Technical Specification, to Facility Operating License NPF-11.

These changes are being submitted for your staff's review and approval to allow operation with Control Rod Drive (CRD) 10-47 withdrawn for the remainder of this cycle.

Attachment A provides background and discussion. The proposed changes are enclosed in Attachment B. The attached change has received both On-Site and Off-Site review and approval. We have reviewed this amendment request and find that no significant hazards consideration exists. Our review is documented in Attachment C.

Commonwealth Edison is notifying the State of Illinois of our request for this amendment by transmitting a copy of this letter and its attachments to the designated State Official.

In accordance with the requirements of 10 CFR 50.170, a fee remittance in the amount of $150.00 is enclosed.

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, H. R. Denton December 9, 1986 Please direct any questions you may have concerning this matter to this office.

Three (3) signed originals and thirty-seven (37) copies of this transmittal and its attachments are provided for your use.

Very truly yours, C. M. Allen Nuclear Licensing Administrator im Attachments A: Background and Discussion B: Technical Specification Change to NPF-18 C: Evaluation of Significant Hazards Consideration D: LaSalle Unit 1 Control Rod Drive (10-47) Uncoupling Safety Evaluation Report dated December 1986 cc: Region III Inspector - LSCS A. Bournia - NRR M. Parker - State of Ill SUBSCRIBED AND SWOR to beforg me this 176 day of N t M u.[+t u , 1986 (Editk Lk l sGW Notary Public 2519K l

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4 ATTACHMENT A TECHNICAL SPECIFICATION CHANGE REQUEST LASALLE COUNTY STATION UNIT 1 BACKGROUND AND DISCUSSION BACKGROUND On November 22, 1986, during the performance of control rod drive (CRD) cxercising on Unit 1, CRD 10-47 became uncoupled (loss of position 48 indication and receipt of the overtravel alarm). CRD 10-47 was inserted to position 44 to recouple. Upon withdrawal, the drive failed to reach position 48 (full out). Subsequent troubleshooting revealed the following cbout the status of CRD 10-47:

- the blade was following the drive, (verified by neutron monitoring instrumentation)

- the drive could not be moved to position 48 or overtravel, but only to a position between 47 and 48 (the overtravel alarm was never received again following the initial uncoupling event, the position 46 indication lit up, but neither the position 48 indicator nor the rod fully withdrawn indicator illuminated), and

- coupling verification could not be performed by the normal method (with the drive at position 48, by demanding a single notch withdrawal and verifying that the drive does not go to the overtravel position).

CRD 10-47 has been inserted (along with its three symmetric drives) to position 00 (full in). This complies with Technical Specificaton 3.1.3.6.

Based on the following discussion, Commonwealth Edison (LaSalle County Nuclear Station) is proposing to operate Unit I with CRD 10-47 withdrawn.

Since CRb 10-47 will not be at position 48 for the remainder of the cycle, an operating license amendment is requested to modify Technical Specifications 3/4.1.3.2, 3/4.1.3.3, 3/4.1.3.4, and 3/4.1.3.6 to allow operation with the inability to verify coupling.

DISCUSSION The drive at core location 10-47 was replaced during the recent Unit I refuel outage. CRD 10-47, which experienced numerous cycles (insert and withdrawals) and coupling verifications since April 1986 (fuel load), had performed normally prior to this event, that is, no overtravel alarms were received until November 22, 1986.

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Bered upon a review of the data, and numerous discussions with site personnel, General Electric (GE) has written a Safety Evaluation Report In this report, GE postulates a scenario which l (Attachment A) on CRD 10-47.

best explains the circumstances that were observed to have happened to CRD I l

10-47, it is believed that the cause of the initial uncoupling was an The uncoupling rod which was located in the wrong position in the CRD spud.

inability to reach position 48 is believed to be caused by a bent spud

)- finger which occurred during the recoupling operation. There is a high j -

scobability that couplina (sufficient coupling) d m exist, however, the

' n: mal method (single notch withdrawal from position 48) of coupling i

v:;rification cannot be performed.

4 CE indicated that uncoupling a drive which is not at position 48 from underneath the vessel is rarely successful (as shown from past field d

experience). Therefore, removal of CRD 10-47 may necessitate lifting the blade for uncoupling purposes with the vessel head removed.

l C:ntrol rod 10-47 is a non-control cell rod which is targeted for position

' 48 during the course of the entire cycle (which has approximately 12 months of operation left). Since the computer code which calculates the core thermal limits requires core syssetry, when control rod 10-47 was inserted to position 00, three symmetric rods were also inserted. Short term (one i

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1. month) operation with four control rods fully inserted will have no adverse impact on the cycle, however, these rods should be withdrawn to their target i

positions (or to position 46 for rod 10-47) as soon as possible to avoid future economic losses.

l Scram performance integrity for an uncoupled or improperly coupled Although rod is i

cddressed in the GE Safety Evaluation Report (Attachment D).

j equipment damage could result (ie - velocity limiter, lock plus, spud),

there is no threat 12 shi 1M11 gr pressure boundary intearity 3DA DE dearadation gf. screm perfomance. CRD 10-47 will be monitored during normal l

< cperation and scram perfomance will be evaluated during unplanned scrams.

Control rod drop accident (RDA) concerns are also addressed in the Gg Safety Evaluation Report (Attachment D). RDAs are a concern at any power level, j

however, the severe consequences, which require Technical Specification The Rod controls (3/4.1.4), occur at only low power (less than 20% power).

Sequence Control System (RSCS) at I.aSalle controls the sequence of rod i

withdrawal at low powers to minimize the severity (limits peak fuol The proposed I enthalples to below 280 cal /sm) should a RDA occur.

! cperational requirements on rod 10-47 would negate RDA concerns by leaving l

the rod inserted and disarmed when less than or equal to 20% power, and verifying control blade following when the drive is withdrawn above 20%

power.

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1 The operational requirements listed below outline the approach to be used when operating with rod 10-47 withdrawn. The basis for such operation is the high probability that coupling does exist. However, if for some i

unforseen reason inadequate coupling exists, there is assurance that there would be no degradation of screm performance and actions can be taken to Cinimize RDA concerns.

- CRD 10-47 will be fully inserted and diserined when less than or equal to 20% power. This negates RDA concerns at low power and satisfies the l

basis of Technical Specification 3/4.1.3.6.

- During withdrawal of CRD 10-47 to its target position (46), neutron l

instrumentation (LPRM or TIP) will be monitored to verify that the control blade is following the drive. This satisfies the basis of Technical Specification 3/4.1.3.6. If neutron instrumentation response does not verify that the rod is following the drive, CRD 10-47 will be inserted to position 00.

-!~ - CRD 10-47 will not be scheduled for surveillance scram testing l

(3/4.1.3.2) until required by the technical specifications. Should

. surveillance scram testing of the drive be required during the course of the cycle, an additional time increment (the scram time between notch positions 03 and 01, located in the drive's buffer region) will be added to the scram times of positions 45, 39, 25, and 05. This additional time increment will conservatively compensate for using position 46, instead of position 48, as the fully withdrawn position.

I For Technical Specifications 3/4.1.3.2, 3/4.1.3.3, and 3/4.1.3.4 the fully withdrawn position for CRD 10-47 is position 46.

' - CRD 10-47 will be exercised weekly (Technical Specification 3/4.1.3.1).

The proposed Unit 1 license amendment (See Attachment B) modifies the

! following:

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- Technical Specifications 3/4.1.3.2, 3/4.1.3.3, and 3/4.1.3.4 are foot-l 4

noted to reflect the fact that control rod 10-47 may use position 46 for the fully withdrawn position, and

- Technical Specification 3/4.1.3.6 is footnoted to reflect the fact that control rod 10-47 is exempt from the coupling check.

l CONCLUSION b

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Based on the previous discussion, Commonwealth Edison (LaSalle County

! Nuclear Station) is proposing to operate Unit I with CRD 10-47 withdrawn j

(but unable to reach position 48). Since the operational requirements placed on CRD 10-47 meet the bases for the Technical Specifications involved i

(3/4.1.3.2, 3/4.1.3.3, 3/4.1.3.4, and 3/4.1.3.6), consonwealth gdison requests NRC review and approval of the proposed Unit 1 operating license

! amendment for Cycle 2.

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