ML17305B118

Application for Amend to License NPF-41,revising Tech Spec Section 4.4.4.3,extending Performance Interval for Steam Generator Eddy Current Exam to End of Present Fuel Cycle

ML17305B118
Person / Time
Site:	Palo Verde
Issue date:	10/06/1990
From:	Conway W ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17305B119	List: ML17305B118 ML17305B120
References
16103526-WFC-JR, NUDOCS 9010230286
Download: ML17305B118 (15)

Text

ACCELERATED DI TRIBUTION DEMONST TION SYSTEM I

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9010230286 DOC.DATE: 90/10/06 NOTARIZED: YES DOCKET FACIL:STN-50-528 Palo Verde Nuclear Station, Unit 1, Arizona Publi 05000528 AUTH. NAME AUTHOR AFFILIATION CONWAY,W.F. Arizona Public Service Co. (formerly Arizona Nuclear Power RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk) R

SUBJECT:

Application for amend to License NPF-41,revising TS Section I 4.4.4.3,extending performance interval for SG eddy current.

COPIES RECEIVED: LTR SIZE: D DISTRIBUTION CODE: A047D ENCL TITLE: OR Submittal: Inservice Inspection/Testing/Relief from AS Code S

05000528 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD5 LA 1 0 PD5 PD 1 1 D PETERSON,S. 2 2 TRAMMELL,C. '2 2 D

INTERNAL: ACRS 6 6 AEOD/DSP/TPAB 1 1 NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB 7E 1 1 NUDOCS-ABSTRACT 1 1 OC/LFMB 1 0 OGC/HDS1 RES/DSIR/EIB 1

1 0

1 E~ 1 1 EXTERNAL EGGG BROWN i B 1 1 EGGG RANSOME,C 1 1 NRC PDR 1 '

NSIC 1 1 NOTES: 1 1 D

A D

D NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM P 1-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 25 ENCL 22

Arizona Public Service Company P.O. BOX 53999 ~ PHOENIX, ARIZONA 85072<999 WILLIAMF. CONWAY 16103526-WFC-JRP EXECUTIVEVICEPRESIDENT NUCLEAR October 06, 1990 Docket No. STN 50-528 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station Pl-37 Washington, D. C. 20555

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Unit 1 Technical Specification Amendment Extension of Interval for SG Eddy Current Examination File: 90-F-005-419.05; 90-056-026 This letter is being provided to request an amendment to the PVNGS Unit 1 Technical Specifications Section 4.4.4.3, Inspection Frequencies (inservice inspection of steam generator tubes). The proposed change would extend the performance interval for the Unit 1 steam generator eddy current examination from the existing intervals of not less than twelve nor more than twenty-four calendar months after the previous inspection to the end of the present fuel cycle (Cycle 3).

The performance interval for the Unit 1 steam generator eddy current examination should be extended to the end of the present fuel cycle to coincide with the next refueling outage presently scheduled to begin February 1, 1992. The bases and justification for the proposed change is presented herewith as an attachment.

Also enclosed with this amendment request are:

A. Description of the Amendment Request B. Purpose of the Technical Specification C. Need for the Technical Specification Amendment D. Basis for No Significant Hazards Consideration E. Safety Analysis of the Proposed Amendment Request F. Environmental Impact Consideration Determination G. Marked-up Technical Specification Change Page(s)

Pursuant to 10 CFR 50.91(b)(1), and by copy of this letter and attachment, we have notified the Arizona Radiation Regulatory Agency of this request for a Technical Specification Amendment.

90i0230286 901006 PDR ADOCK 05000528 P PDC No+7

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U. S. Nuclear Regulatory Commission Attention: Document Control Desk Page 2 Should you have any questions, please call Joseph R. Provasoli oE my staff at (602),,340-4160.

Since'rely, WFC/JRP/j le Attachment cc: S. R. Peterson (all w/attachment)

C. M. Trammell J. B. Martin D. H. Coe C. F. Tedford

STATE OF ARIZONA )

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COUNTY OF MARICOPA )

I, W. F. Conway, represent that I am Executive Vice President Nuclear, that the foregoing document has been signed by me on behalf of Arizona Public Service Company with full authority to do so, that I have read such document and know its contents, and that to the best of my knowledge and belief, the statements made therein are true and correct.

W. F. Conway Sworn To Before Me This ~0 Day Of 1990.

Notary Public My Commission Expires

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ATTACHMENT I

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A. DESCRIPTION OF THE AMENDMENT RE VEST 4

The proposed Technical Specification amendment would revise Section

'.4.4.3','Inspection Fr'equencies (inservice inspection of steam generator tubes) to extend the performance interval for the Unit 1 steam generator eddy current examination. The inspection frequency currently applicable to the Unit, 1 steam generators is as follows, "Subsequent inservice inspections shall be performed at intervals of not less than twelve nor more than twenty-four calendar months after the previous inspections."

The proposed change would extend the performance interval for the Unit 1 steam generator eddy current examination to the end of the present fuel cycle (Cycle 3).

B. PURPOSE OF THE TECHNICAL SPECIFICATION The surveillance requirements for inspections of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained; The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

CD NEED FOR TECHNICAL SPECIFICATION AMENDMENT The Technical Specification surveillance requirements require that subsequent inservice inspections be performed at intervals of not less than twelve nor more than twenty-four calendar months after the previous inspection. The most recent inservice inspection of the Unit 1 steam generator tubes was completed in July 1989 during the second refueling outage.

The Unit 1 steam generator tubes have been eddy current examined on four

'revious occasions. The information presented below lists the time frame of each examination and the number or percentage of steam generator tubes examined.

ATTACHMENT July - August 1981 January, 1987 October, 1987 July, 1989 1st refuelin 2nd refuelin l.

SG 11 100% 3516 2375 100%

SG 12 100% 3496 3750 1008 The types of problems noted in the Unit 1 steam generators can be summarized as follows:

lower eggcrate wear minor dents and dings minor denting around flow distribution baffle possible loose parts indications minor batwing and vertical strip wear minor sludge buildup The observed conditions are primarily associated with mechanical wear as opposed to chemistry or corrosion problems. The mechanical wear is due to vibration associated with normal plant operations. Based on the inservice inspection results the following tube plugging has resulted:

October 1987 J~ul 1989 SG 11 20 12 SG 12 14 7 These tubes were plugged based on a criterion of 30$ through wall indication. The overall results of the examination are encouraging in that after completion of the second cycle, the total number of degraded and/or defective tubes is minimal and no significant wear pattern has been observed.

The performance interval for the Unit 1 steam generator eddy current examination should be extended to the end of the present fuel cycle to coincide with the next refueling outage presently scheduled to begin February 1, 1992. Justification for the extension of the performance interval is as follows:

There are minimal degraded and/or defective tubes in the Unit 1 steam generators.

No significant wear patterns have been observed (based on result from previous inservice inspections).

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ATTACHMENT Continued A 100% eddy current examination of the Unit 1 steam generator tubes was performed,'during the last refueling outage (July 1989). E Adequate, chemistry control was maintained in the Unit 1 steam

'generators, 'during, the shutdown and operational periods to date.

Minimal mechanical wear occurred for the eleven month period following the~ most recent eddy current examination while the unit was shutdown.'nit 1 entered mode 1 on June 30, 1990, ending its second refueling outage which began on March 5, 1989 when the unit entered Mode 3. If the proposed amendment request were not approved by the NRC the Unit 1 steam generators would be eddy current examined during the January 1991 surveillance test outage. Thus, the steam generators would only have been subjected to mechanical wear, which is the most significant contributor to tube degradation, for approximately seven months. As a result, the eddy current examination will not likely be meaningful in assessing tube wear patterns nor will there have been a sufficient period of operation to cause a concern for a significantly increased degradation of the tubes. The eddy current examination would also require mid-loop operation. And finally, there would be increased radiation exposure incurred while installing and removing nozzle dams.

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. A discussion of these standards as they relate to the amendment request follows.

Standard 1 -- Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed Technical Specification amendment will not increase the probability or consequences of an accident previously evaluated. The radiological releases calculated for the steam generator tube rupture event with a loss of offsite power and a fully stuck open atmospheric dump valve

ATTACHMENT Continued limits, thus assuring the integrity of these systems. Additionally, no violation of the fuel thermal limits occurs, since the minimum DNBR remains above the 1.24 value throughout the duration of the event. Therefore, the proposed change will not increase the probability or consequences of an accident previously evaluated.

Standard 2 -- Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed Technical Specification amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The Chapter 15 analysis assumes that the plant is challenged by a steam generator tube rupture that includes additional events and failures beyond those postulated by the NRC Standard Review Plan (SRP) 15.6.3. In addition to the conservative assumptions of the SRP (loss of offsite power, iodine spiking, etc.), this analysis postulates that the operators open an ADV on the affected steam generator and that it both runs to the full open position and sticks full open for the duration of the transient. The results of which are well within the guidelines of 10 CFR Part 100 for any radiological releases and the RCS and secondary system pressures are well below the design pressure limits. Therefore the proposed Technical Specification amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

1 Standard 3 -- Involve a signifi'cant reduction in 'a margin of safety.

The proposed Technical Specification amendment does not involve a significant reduction in a'argin of safety because no changes are being made to the way the facility 'is'being operated. Thus, no new failure modes are being introduced.

'I If a )) I steam. generator 'tube rupture were to occur, diagnosis of the event is facilitated by radiation monitors, which initiate alarms and inform the operator of abnormal levels and that corrective operator action is required. Additional diagnostic information is provided by RCS pressure and pressurizer level response indicating a leak, and by level response in the affected steam generator.

The most limiting steam generator tube rupture event is for a leak flow equivalent to a double-ended rupture of a U-tube at full power conditions.

This event has been analyzed for Palo Verde (UFSAR Section 15.6.3) and concludes that the resultant radiological releases are well within the 10 CFR 100 guidelines and the RCS and secondary system pressures are well below 110% of the design pressure limits and no violation of the fuel thermal limits occurs. Therefore, the proposed Technical Specification amendment will not involve a significant reduction in a margin of safety.

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ATTACHMENT Continued E. SAFETY ANALYSIS OF THE PROPOSED AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Final Safety Analysis Report (UFSAR). The amendment will not make changes to the facility. The additional time between surveillance intervals will not increase the probability of failure of equipment. The actual operating time on the Unit 1 steam generators will be within the twelve to twenty-four month requirement for the surveillance interval.

The Unit 1 steam generator tubes were 100% eddy'urrent examined during the last refueling outage (July, 1989). The overall results after completion of the second fuel cycle are encouraging in that the total number of degraded and/or defective tubes were minimal and no significant wear patterns have been observed.'This is based on the results of three inservice inspections during the first two fuel cycles.) Therefore, the proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or'malfunction of equipment important to safety previously evaluated in the UFSAR.

The proposed Technical Specification amendment will not create the possibility of an accident or malfunction of a different type than any evaluated previously in the UFSAR. No physical changes are being made to the facility nor are there any changes being made which affect the operation of the facility. No new failure modes are being introduced by the change. The UFSAR Chapter 15 analysis includes a steam generator tube rupture with concurrent loss of offsite power and a stuck open ADV. This analysis concludes that radiological releases are within federal guidelines and the RCS and secondary pressures are well below the design pressure limits. Therefore the proposed Technical Specification amendment will not create the possiblity of an accident or malfunction of a different type than any evaluated previously in the UFSAR.

The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for any Technical Specification. During the Unit 1 second refueling outage 100% eddy current examination was performed on both steam generators with no corrosion related defects found.

There has been little or no mechanical wear on the steam generators since the last refueling outage due to the fact of an unusually lengthy outage.

If eddy current examinations were conducted during the January surveillance outage the steam generator will have had only seven months of operations.

ATTACHMENT Continued During the most recent Unit 1 refueling outage adherence to wet lay-up conditions were within specifications except for the period of time that nitrogen overpressure was not maintained to allow for the atmospheric dump valve modifications. Combustion Engineering (CE) was requested to evaluate the long-term corrosion effects on steam generator materials that could be attributed to any out of specification lay-up condition. Based on conclusions from CE, no corrosion mechanism should have been initiated during this refueling outage which would require an eddy current examination prior to the next scheduled refueling. Therefore, the proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for any Technical Specification.

ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed Technical Specification amendment request does not involve an unreviewed environmental question because operation of PVNGS Unit 1 in accordance with this change would not:

(1) Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board, Supplements to the FES, Environmental Impact Appraisals, or in any decisions of the Atomic Safety and Licensing Board; or (2) Result in a significant change in effluents or power levels; or Results in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.

MARKED-UP TECHNICAL SPECIFICATION PAGE (See attached page 3/4 4-'13)

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