ML20204G022

Safety Evaluation Supporting Amend 76 to License DPR-61 Re Operation of Plant Using Interim Acceptance Criteria for Steam Generator Tube Repair.Salp Evaluation Encl

ML20204G022
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	07/30/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20204G018	List: ML20204G012 ML20204G022
References
TAC-67463, NUDOCS 8608070040
Download: ML20204G022 (11)

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1. UNITED STATES

[ .g NUCLEAR REGULATORY COMMISSION l T 5 8 WASHINGTON, D. C. 20666 '

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 76 TO FACILITY OPERATING LICENSE NO. DPR-61 CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT DOCKET NO. 50-213

1.0 INTRODUCTION

Connecticut Yankee Atomic Power Company (CYAPCO), using an improved signal processor, identified 575 steam generator tube ends during the 1986 re-

" fuelling outage as having " undefined eddy current signals" in the roll expansion region.

A tube with an undefined signal was removed during this past outage from Haddam Neck Plant Steam Generator No. 3 to assist in characterization of the undefined signals. Preliminary results of a destructive examination of the removed tube indicated that the undefined signal in that tube was due to inside diameter initiated cracks at the center of the roll expan-

,' sion region. The cracks were 0.25 inches long, oriented 30' from the longitudinal axis and up to 82% through-wall. The cracks were caused by primary water stress corrosion.

Post-outage investigations have been conducted in an attempt to further characterize the undefined signals. A discussion of the post-outage investigation was presented to the NRC Staff by letter dated June 20, 1986 j (Reference 1) which concluded that the potential flaws associated with i these signals did not represent an undue risk to public health and safety.

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1 By letter dated June 26,1986 (Reference 2) the NRC Staff agreed with this conclusion on the basis that there is reasonable assurance that the potential flaws associated with these signals will not result in cata-strophic tube failure in the near term and that any propagation of existing flaws will be revealed by an increased primary-to-secondary

, coolant leakage rate. The letter also stated that the completed evalu-ation should serve as the basis for a license amendment to 'specify the acceptance criteria for the steam generator tubes in the tubeshedt region. The proposed corrective action plan was to be completed by i September 30, 1986 and a proposed license amendment was requested to be submitted by October 31, 1986. The Staff also requested a meeting with the Licensee towards the end of August 1986 to discuss the result of the investigation. .

Subsequent investigations of the undefined signals ruled out the possi-bility of flaws in 35 of the 575 tube ends under consideration. Esti-mated flaw depths for the remaining 540 tube ends with undefined signals ranged from 22% to 100% through-wall. This sizihg estimate assumed the entire undefined signal was due to a flaw, even though the existence of a flaw could not be confirmed in all cases. ,

2.0 DISCUSSION i By letter dated July 24, 1986 (Reference 3) the licensee submitted an analysis to support the proposed interim structural acceptance criteria I

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for flaws in the rolled region of the tubes. Specifically, the proposed structural acceptance criteria for the rolled region of the steam gen-erator tubes are as follows:

I No defect in the uppermost one inch of hard roll; and Any size defect is acceptable below the uppennost one inch.

These interim repair criteria were applied to the tubes which were potentially flawed in the uppermost one inch of the rolled region. A list of 119 tubes was developed that did not satisfy this acceptance criterion and a decision was made to bring the plant to cold shutdown to plug these tubes. The decision was also made to plug Tube 37-73 in Steam 4

Generator No. 2, which is currently in service with a 557, through-wall i

defect (Reference 4).

The Staff has reviewed the proposed interim acceptance criteria for steam t

! generator tubes with flaws in the rolled region of the tube. Based on 1

l this review it is concluded that the application'of these criteria will provide adequate tube integrity for the remainder of the current fuel cycle.

Specifically, the margin of safety on the tube integrity will not be reduced because the interim acceptance criteria were developed in accordance with the intent of Regulatory Guide 1.121 and the Plant Technical Specifications.

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3.0 EVALUATION This evaluation is based on the information provided in the licensee sub-mittal of July 24,1986(Reference 3). The proposed interim acceptance I:

criteria were developed on the basis of considering sufficient frictional force with margin in the rolled region needed to resist the pull out force which the tube might experience during normal and postulated accident conditions.

The preload is primarily due to hard rolling. This preload assumes an elastic-plasticmaterialwithayieldstrengthof37ksi(theminimum measuredvalueforthistubing). The contact pressure obtained was 5457 psi. This contact pressure produces a hoop stress of 37 ksi. For an elastic-plastic material, this contact pressure would be required to stop the radial strain during rolling. After the roller is removed, the tubesheet rebounds elastica 11y to produce a 37 ksi compressive hoop stress.

Thus, the post rolling preload is also 5457 psi.

The rolled tube section was checked for thermal ratchetting. Again, the material was conservatively assumed to be elastic-plastic. Since the externally applied mechanical load produces yield strength of 37 ksi, the thermal stress would have to be greater than 37 ksi to produce ratchetting (Reference 5). The maximum thermal strain was found to be 0.04% which is well below the yield strain of 0.2%. Thus, no ratchetting is likely to occur.

For the material under consideration, there is a themally induced loss of cold preload (i.e., (0.2-0.04)/0.2 = 80%). This preload is restored on each heatup so that the hot preload is equal to the pre-shakedown cold preload. Thus, the 5457 psi contact pressure is obtained on each heatup.

An estimate was made of the relaxation in stress due to creep. Available data at temperatures of 1000* - 1500*F (Reference 6) was extrapolated to 600'F. Based on these data a relaxation of 80 percent was ' estimated over a 35 year period. Thus, the post creep relaxation pressure was detemined

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to be 4342 psi.

Environmental effects were also considered. Boric acid attack on the 1

tubesheet was reviewed and found to be acceptable due to the generally low oxygen concentration and the lack of dissociation at opera, ting temperatures. Further, the corrosion products are insoluble and have a lower density than the parent material. This would tend to tighten the joint:and cut off further access to liquid. Based on experience from fossil plants and lines at 600'F in water environments, no creep damage is considered likely. -

A friction coefficient of 0.18 was used to obtain the magnitude of frictional force (Reference 7). Since the value is for a sliding, greasy nickel on steel interface, it represents a conservative static value.

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The total contact pressure after taking the various factors discussed above into consideration was determined to be 5717 psi for nomal operating conditions and 6842 psi for the Main Streamline Break (MSLB) faulted conditions. This results in required engagement lengths of 0.26 inches and 0.49 inches for the nomal and faulted conditions, respectively. ,

Thus by providing a minimum required sound roll of 1.0 inch, a factor of

. safety of greater than 3.0 for normal operating conditions'and 1.428 or more for faulted conditions have been provided. These factors of safety c ..

a're Jn accordance with Regulatory Guide 1.121 and are in addition to an allowance of 0.1 inches for eddy current measurement errors. The staff, therefore, finds them acceptable.

Due to the current concerns in the areas of inspection accuracy, leakage rates, and residual stresses, it was decided that no defects would be pemitted within the uppemost one inch of roll. For defects below the I

one-inch roll, leakage was considered. Using test data for one-inch rolls

! (Reference 8)', a conservative maximum leak rate was obtained. The maximum l leakage was found to occur during LOCA conditions when the bowing of the l tubesheet dilates the holes in the rolled expansion region. The total secondary-to-primary leak rate for 575 such rolled joints was determined j to be 0.03 gpm. Since this is well below the limits of the technical ,

i l specifications (0.4 gpm), it is considered acceptable.

l Postulated failure modes below the one-inch roll were reviewed to determine their impact, if any, on the roll. Various configurations of cracks were considered (single, two asymmetric, multiple synnetric). It was found that two axial cracks separated by a thin (0.15 inch) ligament would

' punch in' at a lower critical pressure than other configurations. The result of such a ' punch in' was then evaluated to determine whether this might propagate by buckling into the sound roll. It was found that the critical pressure to collapse the sound roll even after a ' punch in' below it was in excess of 1800 psi (1.8 times secondary design pressure). Thus, even limiting postulated situations will not collapse the sound roll due to defects below it.

4.0 CONCLUSION

The staff has reviewed the proposed interim repair criteria for steam 1

generator tubes with flaws in the rolled region of the tube. Based on the review it is concluded that one inch of sound roll has the required .

frictional re'istance s to withstand normal operating, test, and postulated accident loads. A factor of safety of greater than 3.0 under nonnal operating conditions and 1.428 or more under faulted conditions to with-stand the pull out force have been provided. These factors of safety are in accordance with the intent of Regulatory Guide 1.121 and are in addition ,

to an allowance of 0.1 inches for eddy current measurement errors. -

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l Based on a review of,the analysis fer the postulated primary-to-secondary leakage past the sound rolled region, it is concluded that the maximum calculated leakage for all tubes with flaws which may leak will not exceed the limit of 150 gallons per day per steam generator allowed in the Haddam Neck Plant Technical Specifications for design basis accidents.

The staff, therefore, finds the proposed interim repair criteria acceptable for the remainder of the current fuel cycle.

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References

1. J. F. Opeka letter to C. I. Grimes, dated June 20, 1986,

Subject:

Informational Letter on Undefined Signals Observed During Steam Generator Eddy Current Inspection

2. C. I. Grimes letter to J. F. Opeka, dated June 26, 1986,

Subject:

Undefined Signals Observed During Steam Generator Eddy Current Inspection. l

3. J. F. Opeka letter to C. I. Grimes (NRC) dated July 24, 1986,

Subject:

Interim Acceptance Criteria for Steam Generator Tubes With Flaws in Tubesheet Roll Region.

4. J. F. Opeka letter to C. I. Grimes, dated May 9,1986,

Subject:

L Proposed Revision to Technical Specifications Steam Generators and License Amendment No. 76 transmitted by C. I. Grimes letter to J. F. Opeka dated May 14, 1986,

Subject:

Steam Generator Tube Plugging.

5. " Creep Analysis," H. Kraus, John Wiley & Sons, New York 1980, page 166.

6. " Nuclear Systems Handbook," Combustion Engineering Inc., Part 1, Group 4, Section 3. Inconel Alloy 600' Rev. O, November 15, 1972.

7. " Marks Standard Handbook for Mechanical Engineers," Baumeister et al, Eighth Edition. McGraw-Hill, New York, 1978, p. 3-26.

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8. Westinghouse Report WCAP-10267 (Millstone 2 Docket) Figure 6.1-5.

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SALP REPORT PLANT: Haddam Neck Plant LICENSEE: Connecticut Yankee Atomic Power Co.

DOCKET NO.: 50-213 REVIEWER: J. Rajan LICENSING ACTIVITY: Repair of Steam Generator Tubes in the tubesheet region.

i EVALUATION CRITERIA RATING REMARKS

1. Management Involvement 2 and Control in Assuring Quality

2. Apriroach to Resolution 2 of Technical Issues from a Safety Standpoint

3. Responsiveness to NRC 2 Some of the responses Initiatives to NRC concerns lacked sufficient detail and background information, p 4. Enforcement History NA

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5. Reporting and Analysis NA j of Reportable Events y 6. Staffing NA The sununary SALP rating for this submittal is 2. -

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July 30, 1986 The staff has reviewed the information and analyses provided by CYAPC0 in support of the interim acceptance criteria and concludes that one inch of sound roll will provide the required frictional resistance to withstand normal operating, test, and postulated accident loads. The safety factor to withstand the pull out force is greater than 3.0 under nonnal operating conditions and 1.428 or more under faulted conditions. These factors of safety are in accordance with the intent of Regulatory Guide 1.121 and are in addition to an allowance of 0.1 inches for eddy current measurement errors. Additionally, the staff has concluded that the maximum calculated leakage for all tubes with flaws j which may leak will not exceed the Haddam Neck Technical Specification limit for design basis accidents (150 gallons per day per steam generator).

Therefore, we conclude that there is reasonable assurance that the potential flaws in the rolled region of the sub.iect 540 steam generator tubes will not result in catastrophic failure in the near tenn and that any propagation of existing flaws will be revealed by an increased primary-to-secondary coolant leakage rate, which is limited by plant technical specifications. On this basis, tiie staff concludes that operation of the Haddam Neck Plant using the interim acceptance criteria for steam generator tube repair does not present an undue risk to public health and safety for the remainder of the corrent fuel cycle and is, therefore, acceptable. A copy of our Safety Evaluation j concerning the interim acceptance criteria is enclosed.

We believe that it is appropriate for you to continue and complete the evaluation of the undefined signals, which will serve as the basis for a future license amendment, to specify the final acceptance criteria for the roll expansion i region of the steam generator tubes.

As previously stated in our June 26, 1986 letter, we require that your study regarding the SG tube ends, with a proposed corrective action plan, be completed no later than September 30, 1986. A license amendment, with the appropriate

acceptance criteria for the tube ends, should subsequently be submitted by l

October 31, 1986. We again recommend that a meeting be held with the NRC staff i toward the end of August 1986, to discuss your preliminary results regarding l the significance o,f the undefined eddy current signals.

l Sincerely,

_ Original signed by Michael L. Boyle for Christopher I. Grimes Christopher I. Grimes, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing - B '

DISTRIBUTION

Enclosure:

Docket File PAnderson JPartlow As Stated NRC PDR FAkstulewicz BGrimes Local PDR ACRS (10) ISAP Reading cc: See Next Page CGrimes FMiraglia EJordan CCheng WE6a 09p ,NThompson SAP:DPL-B I -B EB: h h

.FAkstulewicz:mn d erso CCheng p()ISAP:DPL-B Grimes 1 /30 /86 1 /30/86 gy6 1/g /86

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