ML20235S550

Safety Evaluation Supporting Amend 96 to License DPR-61

ML20235S550
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	09/25/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20235S503	List: ML20235S499 ML20235S525 ML20235S550
References
TAC-67463, NUDOCS 8710090038
Download: ML20235S550 (4)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIO$

SUPPORTING AMENDMENT NO. 96 TO FACILITY OPERATING LICENSE N0. DPR-61 1 l

CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT I

DOCKET N0. 50-213 j 1

1.0 INTRODUCTION

By letter dated June 1,1987, the Connecticut Yankee Atomic Power Company (CYAPC0) submitted a request for changes to the Haddam Neck Plant Technical Specifications.

The license amendment revises Technical Specifications Section 4.10.0.1 to )

provide long-term acceptance criteria for the steam generator tubes with t defects in the rolled region (bottom four inches of the tube) and to update the bases for this criteria. These changes will not affect repair ' criteria l for flaw indications located outside of the roll expansion region. This  !

license amendment will also modify the current requirement that "The i plugging limit for sleeves will be determined prior to the 1987 refueling outage for Cycle 15," to, "The plugging limit for sleeves will be determined prior to the first refueling outage following sleeve installation," since, to date, no sleeves have been installed at the Haddam Neck Plant. This license amendment also revises Technical Specifications Section 4.10.D.2 to j delete the exclusion of tube row 37, column 73 in Steam Generator #2 from  ;

plugging, since this tube has subsequently been plugged during a mid-cycle {

shutdown in July, 1986. l l

2.0 BACKGROUND

In a conference call on June 20, 1986, CYAPC0 informed the NRC headquarters staff and Region I that, using advanced eddy current inspection equipment, 575 steam generator (SG) tube ends were identified during the 1986 refueling as having " undefined eddy current signals" in the roll expansion region. i CYAPC0 performed an analysis to determine interim structural acceptance criteria assuming flaws in the rolled region of the tubes using the guidance in Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes." The proposed interim acceptance criteria includes ,

conservative allowance for dynamic effects during accidents and uses safety ,

factors consistent with those described in the Regulatory Guide. On the .

bases of its analyses, CYAPC0 concluded that one inch (F* zone) of sound roll has the required frictional resistance for a SG tube to withstand 8710090038 B70925 3 DR ADOCK 0500

normal operating, test and postulated accident loads. Therefore, the l interim acceptance criteria for the steem generator tube at the Haddam Neck Plant does not permit any defects in the uppermost one inch of the rolled region at the bottom of the tube sheet. i The staff reviewed the information and analyses provided by CYAPC0 in support of the interim acceptance criteria and concluded in a Safety Evaluation transmitted by letter dated July 30, 1986, that one inch of sound roll would 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 normal operating I conditions and 1.428 or more under faulted conditions. These factors of l safety are in accordance with the intent of Regulatory Guide 1.121 and are l in addition to an allowance of 0.1 inches for eddy current measurement j errors.

3.0 EVALUATION By letter dated June 1, 1987, CYAPC0 requested a license amendment to provide a long-term plugging criteria for the rolled region of steam 1 generator tubes and provided additional information concerning the bases for the proposed plugging criteria.

By letter dated August 27, 1987, CYAPC0 provided additional information in support of the proposed technical specifications to provide additional assurance that the flaws in the rolled region within the top 1 inch are identified by the current steam generator inspection method including an allowance for errors inherent in the existing eddy current testing methods.

CYAPC0 stated that future inspections of the rolled region will consist of  ;

at least 3 percent of the sample of unaffected tubes plus all degraded tubes identified from previous inspections.

In addition, future SG tube inspection programs at the Haddam Neck Plant will continue to include, but not be limited to, the following practices:

1) The use of a special probe for optimum elevation determination of flaws  !

relative to the F* zone. .

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2) The use of independent data analysis, and conservative resolution of discrepancies by an experienced analyst.

3) Any distorted or questionable signals will be inspected with the rotating pancake coil (RPC). If any tube contains a questionable signal in the F* zone which is not positively resolved by the bobbin-probe or RPC-probe inspections, it will be plugged as a con- ,

servative corrective measure. ,

4) The performance of special inspections and examinations, including tube removal / destruction examination, as appropriate, to ensure that the nature and mechanism of the tube degradation process is accurately defined.

5) The application of eddy current testing equipment technology, as appropriate, for defining tube degradation conditions known to exist in the Haddam Neck Plant SGs or similar PWR SGs.

The staff has reviewed the information contained in the letters dated June 1, 1987 and August 27, 1987 and ccncludes that the interim F* zone found acceptable in the safety evaluation (attached) dated July 30, 1986, is still valid and is acceptable for use as the long term plugging criteria.

Further, the staff concludes that the licensee's ongoing inspection program provides a reasonable level of assurance that defects or flaws are identified and plugged, if necessary.

Therefore, the staff concludes that there is reasonable assurance 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 and that the ongoing licensee inspection program should detect any degradation in the roll expansion region of the tube sheet. On this basis, the staff concludes that operation of the Haddam Neck Plant using the long-term plugging criteria for steam generator tube repair does not present'an undue risk to public health and safety and is, therefore, acceptable.

The proposed changes involving the determination of a plugging limit for sleeved steam generator tubes and the deletion of an exclusion permitting operation with a defective tube (Amendment 76, dated May 24,1986),which has since been plugged, are administrative in nature and are acceptable.

4.0 ENVIRONMENTAL CONSIDERATION

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This amendment involves a change to a requirement with respect te the  !

installation or use of facility components located within the restricted j area as defined in 10 CFR Part 20 and changes to the surveillance require- i ments. The staff has determined that the amendment involves no significant  !

increase in the amounts, and j effluents that may be release,no significant d offsite and that change there isinnot thesignificant types, of any  ;

increase in individual or cumulative occupational radiation exposure. The l Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public  ;

comment on such finding. Accordingly, this amendment meets the eligibility  !

criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant I to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.  ;

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5.0 CONCLUSION

The staff'has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the will not be endangered by operation in the proposed manner,such and (2) public activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0. ACKNOWLEDGEMENT Principal Contributors: F. Akstulewicz, PDISA, NRR and H. Conrad, MTEB,NRR Dated: September 25, 1987 l

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\*****J' July 30, 1986 Docket No.: 50-213 Nr. John F. Opeka, Senior Vice President Nuclear Engineering and Operations

/ Northeast Nuclear Energy Company Post Office Box 270 Hartford, Connecticut 06141-0270

Dear Mr. Opeka:

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SUBJECT:

TUBE PLUGGING CRITERIA FOLLOWING DISCOVERY OF UNDEFINED SIGNALS DURING STEAM GENERATOR EDDY CURRENT INSPECTION OF TUBES IN THE TUBESHEET REGION Re: Haddam Neck Plant

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l In a conference call on June 20, 1986, the Connecticut Yankee Atomic Power Company (CYAPCO) infonned the NRC headquarters staff and Region I that, using advanced eddy current inspection equipment, 575 steam generator (SG) tube ends were identified during the 1986 refueling as having " undefined eddy current signals" in the roll expansion region. CYAPC0 stated that no actions were taken during the outage, as the nature of the signals is unknown and characterization of the anomalies could not be performed. By letter dated June 26, 1986, the NRC stated that it was appropriate for CYAPC0 to complete an evaluation which will serve as the basis for a license amendment to specify the acceptance criteria {

for the roll expansion region of the steam generator tubes.

By your letter dated July 24, 1986, CYAPC0 stated that subsequent investigations j of the undefined signals ruled out the possibility of flaws in 35 of the 575 tube '

ends under consideration. Estimated flaw depths for the remaining 540 tube ends with undefined signals ranged from 22% to 100% through-wall- The sizing 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.

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CYAPC0 performed an analysis to determine interim structural acceptance criteria assuming flaws in the rolled region of the tubes using the guidance in Regulatory I Guide 1.121, " Bases for' Plugging Degraded PWR Steam Generator Tubes." The i proposed interim acceptance criteria includes conservative allowances for l dynamic effects during accidents and uses safety factors consistent with those ' described in the Regulatory Guide. On:the bases of its analyses CYAPC0 concluded that.one inch of sound roll has the required frictional resistance .

for a SG tube to withstand normal operating, test 'and postulated accident loads. .

Therefore, the interim acceptance criteria for the steam generator tube ends -

at the Haddam Neck Plant does not permit any defects in the uppennost one inch of the rolled region of the steam generator tube. CYAPC0 has identified a total of 119' tubes containing eddy current indications which did not meet the interim acceptance criteria and has plugged these tubes during the recent plant ,

shutdown. I n u nmm D P V U \,J (v ms a v

July 30, 1986

-2 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 nonnal 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 ir.ches for eddy current measurement errors. Additionally, the staff has concluded that the maximum calculated leakage for all tubes with flaws 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 subject 540 steam generator tubes will not result in catastrophic failure in the near term and that any propagation of existing flaws will be revealed by an increased primary-to-secondary coolant j

leakage rate, which is limited by plant technical specifications. On this basis, the staff concludes that operation of the Haddam Neck Plant using the interim acceptance criteria for steam generator tube repair does not present l an undue risk to public health and safety for the remainder of the current l fuel cycle and is, therefore, acceptable. A copy of our Safety Evaluation I concerning the interim acceptance criteria is enclosed. {

1 We believe that it is appropriate for you to continue and complete the evaluation l 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 region of the steam generator tubes.

As previously stated in our June 26, 1986 letter, we require that y'our 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 October 31, 1986. We again reconnend that a meeting be held with the NRC staff toward the end of August 1986, to discuss your' preliminary results regarding the significance of the undefined eddy current signals.

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Sincerely, Original signed by Michael L. Boyle for Christopher 1. Grimes Christopher I. Grimes, Director Integrated Safety Assessment . j Project Directorate  !

Division of PWR Licensing - B -

DISTRIBUTION

Enclosure:

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$GM 4g- NThompson SAP:DPL-B I -8 EB:pj AIerso FAkstulewic2:mn 9 /20'/86 4 12mo' nCChen ' "gN ((ISAP:DPL-8 Grimes

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1 Nr. John F. Opeka Connecticut Yankee Atomic Power Company Haddam Neck Plant i 1

cc:

Gerald Garfield, Esquire Kevin McCarthy, Director Day, Berry. A Moward Radiation Control Unit Counselors ! Law Department of Environmental City Place Protection i

Hartford, Connecticut 06103-3499 State Office Building Hartford, Connecticut 06106 Haddam Neck Plant Richard H. Kacich, Supervisor RDF #1 Operating Nuclear Plant Licensing Post Office Box 127E Northeast Utilities Service Company East Hampton, Connecticut 06424 Post Office Box 270 Hartford, Connecticut 06141-0270 Edward J. Mroczka Vice President, Nuclear Operations Northeast Utilities Service Company Post Office Box 270 i Hartford, Connecticut 06141-0270 Board of Selectmen Town Hall Haddam, Connecticut 06103 State of Connecticut Office of Policy and Management l ATTN: Under Secretary Energy Division 1 80 Washington Street ,

Hartford, Connecticut 06106 Resident Inspector Haddam Neck Nuclear Power Station i c/o U.S. NRC East Haddam Post Office -

East Haddam, Connecticut 06423

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Regional Administrator, Region I '

U.S. Nuclear Regulatory Comission 631 Park Avenue King of Prussia, Pennsylvania 19406 i l

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I SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 76 TO FACILITY OPERATING LICENSE RG"'T>TR-61 CONNECTICUT YANKEE AT6MIC _ POWER COMPANY HADDAM HECK PLANT DOCKET NO. 50-213 1

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 octage from Haddam Neck Plant steam Generator No. 3 to assist in characterization of the undefined signals. Preliminary results of a destructive examination I 1

of the removed tube indicated that the undefined signal in that tube was

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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 I

longitudinal axis and up to 82% through-wall. The cracks were caused by

. t, primary water stress corrosion.

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1 Post-outage investigations have been conducted in an attempt'to further characterize the undefined signals. A discussion of the post-outage

. 1 investigation was presented to the NRC Staff by letter dated June 20, 1986 l '

(Reference 1) which concluded that the potential flaws associated with these signals did not repre';ent en undue risk to public health and safety.

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. j i l 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-

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strophic tube failure in the near term and that any propagation of existing flaws wi11 be revealed by an increased primary-to-secondary

'f coolant leakage rate. The letter also stated that the completed evalu- j ation should serve as the basis- for a license amendment to 'specify l

l the acceptance criteria for the steam generator tubes in the tubesheet '

region. The proposed corrective action plan was to be completed by September 30, 1986 and a proposed license amendment was requested to be submitted by October 31, 1986. The Staff also requested a meeting with I 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 ,

i ranged from 22% to 100% through-wall. This sizing estimate assumed the entire undefined signal was due to a flaw, even though the existence of a flaw could not be confinned in all cases.

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

for fl.aws 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:

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

These interim repair criteria were applied to the tubes which were potentially flawad in the uppennost one inch of the rolled regio,n. 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 Pl ug these tubes. The decision was also made to plug Tube 37-73 in Steam Generator No. 2, which is currently in service with a 55% thro, ugh-wall defect (Reference 4).

The Staff has reviewed the proposed interim acceptance criteria for steam generator tub'es with flaws in the rolled region of the tube. Based on this review it is concluded that the application'of these criteria.will prdvide 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 l.121 and the Plant Technical Specifications.

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9 3.0 EVALUATION This evaluation is based on the infomation provided in the licensee sub-mittal of July 24,1986 (Reference 3). The proposed interim acceptance 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-plastic material with a yield strength of 37 ksi (the minimum measured value for this tubing). 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 elastically to produce a 37 ksi compressive hoop stress. ]

j Thus, the post rolling preload is also 5457 psi.

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The rolled tube section was checked for thermal ratchetting. Again, the

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material was conservatively assumed to be elastic-plastic. Since the externally applied mechanical load produces yield strength of 37 ksi, the

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thermal stress would have to be greater than 37 ksi to produce ratchetting (Reference 5). The maximum themal strain was found to be 0.04% which is l well below the yield strain of 0.2%. Thus, no ratchetting is likely to occur.

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For the material under consideration, there is a thennally 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 determined

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to be 4342 psi.

Environmental effects were also considered. Boric acid attack on the tubesheet was reviewed and found to be acceptable due to the generally low oxygen concentration and the lack of dissociation at operating 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 normal j

operating conditions and 6842 psi for the Main Streamline Break (MSLB) f faulted conditions. This results in required engagement lengths of. 0.26 inches and 0.49 inches for the normal 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 nonnal operating conditions and 1.428 or more for faulted conditions have been provided. These factors of safety a're'Tn 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 permitted within the uppennost one inch of roll. For defects below the 1

one-inch roll, leakage was considered. Using test data for one-inch rolls 1

(Reference 8)', a conservative maximum leak rate was obtained. The maximum j J

leakage was found to occur during 'LOCA conditions when the bowing of the tubesheet. dilates the holes in the rolled expansion region. The total secondary-to-primary leak rate for 575 such rolled joints was determined to be 0.03 gpm. Since this is well below the limits of the technical specifications (0.4 gpm), it is considered acceptable.

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Postul_ated failure modes below the one-inch roll were reviewed to determine -

l their impact, if any, on the roll. Various configurations of cracks were 1 considered (single, two asymetric, multiple symetric). 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 l.

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 l

to defects below it. ]

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4.0 CONCLUSION

The staff has reviewed the proposed interim repair criteria for steam j generator tubes with flaws in the rolled region of the tube. Based on the i

review it is concluded that one inch of sound roll has the required l frictional re'sistance to withstand nonnal operating, test, and postulated accident loads. A factor of safety of greater than 3.0 under nonnal .

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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 l 1

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. i 1

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Based on a review of the analysis for 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:

l Undefined Signals Observed During Steam Generator Eddy Current 1 i

Inspection.

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

Subject:

1 Interim Acceptance Criteria for Steam Generator Tubes With Flaws in a l

Tubesheet Roll Region.

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

Subject:

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 Handboo'k for Mechanical Engineers," Baumeister.et al, Eighth Edition, McGraw-Hill, New York,1978, p. 3-26.

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

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