B14087, Proposed Tech Specs Re Steam Generator Repair Criteria

Proposed Tech Specs Re Steam Generator Repair Criteria

ML20141M348
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Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	07/31/1992
From:	CONNECTICUT YANKEE ATOMIC POWER CO.
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ML20141M345	List: B14087, Proposed Tech Specs Re Steam Generator Repair Criteria ML20141M344
References
B14087, NUDOCS 9208110374
Download: ML20141M348 (11)

Text

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t Docket No. 50-213 -

B14087 .

i Attachment 2 Haddam Neck Plant Proposed Revision to Technical Specifications July 1992 9208110374 920731 ,

PDR ADOCK 05000212

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4 REACTOR COOLANT SYSTEM SURVElllANCE R "VIREMENTS (Continued)

1) All nonplugged degraded tubes.

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2) Tubes in thosa areas where experience has indicated potential problems, and '

3) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspecticn, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

If any tube does not permit the passage of a 0.460 inch probe, this tube shall be plugged.

c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:

1) The tubes selected for these samples - include the tubes from those areas of the tube sheet array where- tubes with imperfections were previously found, and The inspections include those portions of the tubes there 2) imperfections were previously found.

The results of each sample inspection shall be classified into one of the following three categories:

{Attgary Inspection Results C-1 Less than 5% of the total tubes inspected are d6 graded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected _are defective, or between _5% and 10% of-the total tubes inspected are degraded tubes._

C-3 More- than 10% of the total tubes inspected are degraded tubes or more than 1% of' the inspectect tubes are defective.

NOTE: In all inspections, previously degraded tubes must _ exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations. This does not apply within the tube-to-tubesheet roll region.

HADDAM NECK 3/4 4-23 i

0086 Amendment No. J U

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REACTOR COOLANT SYSTEM SURVElllANCE RE0VIREMENTS (Continued) i 4.4.5.4 Mcentance Criteria

a. As used in this specification:

1) LmoerfectiDD means an exception to the dimensions, finish of contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing i

indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections;

2)

Dearadation means a service-induced cracking,

wastage, pitting, wear or general corrosion occurring on either inside or outside of a tube;

3) Sound Roll means the expanded portion of the tube which is 1 free of imperfections; I

4) Dearaded Tube means a tube containing imperfections l greater than or equal to 20% of the nominal wall thickness caused' by degradation above the tubes roll expansion region. Also a tube with an imperfection of any depth in  !

the region between the top of the roll expansion and one-half inch below the uppermost one inch of sound roll is considered a degraded tube;

5)  % Dearadation means the percentage of the tube wall I thickness affected or removed by degradation; l 6) Defect means an imperfection of such severity that it

exceeds the plugging limit. A- tube or sleeve containing l a defect is defective; t

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)) Pluaaina limit means the imperfection depth at or beyond I which the tube or sleeve shall be removed from service.

The tube plugging limit shall be equal to 50% of the nominal tube wall thickness for tubes.*

For the roll expansion region including the transition I region between the expanded-and unexpanded portions of the tube (bottom five inches of the tube) the following criteria apply:

a) Any imperfection is acceptable provided there is 1-inch of sound mil above the imperfection, b) Any tube containing an imperfection of any depth which does not have one inch of sound roll above the imperfection shall - be repaired or removed from service, unless the following criteria can be met:

H/DDAM NECK o t, 7 3/4 4-25

4 REACTOR COOLANT SYSTEM SVRVEILLANCE RE0VIREMENTS (Continued)

2) The imperfection can be characterized as a crack which is primarily axial in orientation (axial extent greater than or equal to the circumferential extent). Tubes with identified circumferential1y oriented cracks should be repaired or removed from services.

2) The crack is less than or equal to 0.45 inches  !

in axial length. '

3) Tubes with arrays of axial cracks may be retained in service provided the cumulative extent of circumferential cracking, which theoretically could be present between the axial cracks and remain undetected by the nondestructive inspection technique utilized,

, does not exceed 1 inch of the tube circumference.

4) The number of tubes with characterized axial cracks retained in service is limited such that the dose contribution from the aggregate tube leakage will be limited to a small fraction of 10CFR100 dose guideline values in the event of a steam line break.-

• The plugging limit for sleeves will be determined prior to the first refueling outage following sleeve installation.

1 HADDAM NECK 3/4 4-254 0081

sp_cJpR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) t 7) Unserviceable describes the condition of a tube if it contains j i a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified -

in Specification 4.4.5.3c., above;

8) Tube Inspection means an inspection of the stea'm generator tube from the hot leg entry point completely around the U-bend to the top support of the cold leg; or an inspection from the point of entry (hot leg or cold leg) completely 4 around the U-bend to the opposite end.

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or sleeve

• all tube:. exceeding the plugging limit as defined in Specification 4.4.5.4.a.6 and plug all defective sleeves) required by Table 4.4-2.

4.4.5.5 Reports

a. Following the completion of each inservice inspection of steam generator tubes, a Special Report documenting the number of tubes plugged, sleeved or dispositioned using the criteria defined in Specification 4.4.5.4.a.7, in each steam generator shall be reported to the Commission within 15 days pursuant to Specification 6.9.2; 4 b. The complete results of the steam generator tube inservice a' inspection shall be submitted to the Commission in a Special Report pursusnt to Specification 6.9.2 within 90 days following the complecion of the inspection. This Sr.ecial Report shall include:

1) Number and extent of tubes inspected,

2) Location and percent of wall-thickness penetration for each indication of an imperfection, and

3) Identification of tubes plugged, sleeved, or dispositioned l using the criteria defined in Specification 4.4,5.4.a.7 I

c. Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 prior to resumption of plant operation. This report shall provide a description- of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

Tube sleeving shall be performed in accordance with the Connecticut fankee Steam Generator Sleeving Report, Ravision 1, transmitted by .W/o WCAP-11009.WCAP-11008 Withheld (Ref 10CFR2.790)|letter dated January 7,1986]].

HADDAM NECK 3/4 4-26 cos7 AmendmentNo.f5

REAC10R COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continued)

. over a large area. Wastage is easily detectable during inservice inspec-tion, so any flaws of this form will be detected during the scheduled refueling outage inspections. Plugging or sleeving is required for any tubes with defects that have penetrated through 50% or more of the tube-wall thickness. Since wastage is a very slow corrosion process, unplugged tubes or sleeves with less than 50% through-wall defects do not pose a safety p roblem. Of course, these tubes are inspected each outage to measure their defect size, and they are plugged or repaired if necessary.

Another form of corrosion is denting, which is caused by the rapid produc-tion of iron oxide within the tube / support crevice region. As this oxide is produced, it fills the gap between the tube / support structure subsequently pushing with sufficient force on the tube to cause a dent. This dent causes i stresses which can lead to stress corrosion cracking of the tube. Those i tubes which do not permit the passage of a 0.460 inch diameter probe are l plugged.

Stress corrosion cracking is caused by a combination of stress in the tube I with or without coincident adverse chemistry. Once a stress corrosion crack begins to form, it may propagate rapidly. If_ this occurs, a through-wall crack may develop during operation and cause primary-to-secondary leakage.

. To limit the extent of tube leakage during operation, the primary-to-secondary leakage has a limit of 150 gallons per day per steam generator as defined in Section 3.4.6.2c. Cracks having a primary-to-secondary !aakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated

accidents. Operating plants have demonstrated that primary-to-secondary leakage of 150 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown or air ejector exhaust.

Leakage in excess of this limit will require plant shutdown and an unsched-uled inspection, during which the leaking tubes will be located and plugged or sleeved as dafined in Specification 4.4.5.4.a.7. l Pitting is another form of corrosion. Pits are small, pinhole-type defects which are caused by chemical impurities adhering to the tube. Pits can corrode the tubes faster than wastage, but because of their size, they have very little effect on tube integrity. Pits can be detected during inservice

inspections, and are plugged or repaired if the defect size exceeds 50%.

Also, leakage caused by pitting 'uring operation would be monitored and measured as discusled above. - Te .ag has shown that pits will cause tube leaks before affecting tube integrity.

If a defect should develop in service, it would be found during scheduh inservice steam generator tube examinations. Plugging or sleeving will be required for all tubes with imperfections equal to or exceeding the plugging limit as defined in Soecification 4.4.5.4.a.7. Tubes containing sleeves with l 1mserfections exceedN +he plugging limit will be plugged. Steam generator tuae inspections' of gr ating plants have demonstrated the capability to i

reliably detect degradamn that has penetrated 20% of the original tube wall thickness and to axially locate the degradation in the rolled region with an accuracy of +0.1 inch.

HADDAM NECK ~ B 3/4 4-4 0088

4 REACTOR COOLANT SYSTEM BASES 3/4.4.5,JTEAM GENERATORS (Continued)

The plugging or sleeving limit defined in Specification 4.4.5.4.a.7 is based I on the requirements set forth in Regulatory Guide 1.121. These requirements are also the bases for demonstrating that a tube imperfection is acceptable

regardless of its depth provided it is located below one inch of so"nd roll or is characterized as a crack in the roll expansion region which .aeets the i requirements defined in Specification 4.4.5.4.a.7 b. The number o'f tubes with characterized cracks retained in service is limited to such that the dose contribution from the aggregate tube leakage will be limited to a small fraction nf 10CFR Part 100 dose guideline values in the event of a steau line break. Using conservative assumptions, a postulated post accident leak rate of 100 gpm would result in doses .less than 10% of 10CFR100 limits. Ten l I

percent is considered a small fraction of 10 CFR 100 limits in this instance.

Whenever the results of any steam generator tubing inservice inspection fall .

into Category C-3, these results will be reported to the Commission in a l Special Report pursuant to' Specification 6.9.2 within 30 days and prior to resumption of plant operation. Such cases will be considered by the Commis-sion on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and

, revision of the Technical Specifications, if necessary.

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 1 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provid-

< ed to monitor and detect leakage from the reactor coolant pressure boundary, i

This technical specification ensures a reliable means of detecting unidenti-fled leakage in the reactor coolant system which potentially could be due to-

! a circumferential through-wall flaw in primary system piping. The required instrument sensitivity is 1 gallon per minute in -4 hours as stated in Condition (2) of Generic Letter 84-04. Because the Volume Control Tank level Monitoring System and the Contair. ment Main Sump Level (Harrow Range)

Monitoring System are not seismically qualified, surveillance requirements for a seismic event greater than one half the Safe Shutdown Earthquake (SSE) are imposed.

The RCS Leakage Detection Systems required by this specification are.provid-ed to monitor and detect leakage from the reactor coolant pressure boundary.

These Detection Systems are consistent with the recommendations of Regula-

. tory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE B0UNDARY LEAYAGE of any magnitude i ' acceptable since it may be indicative of an impending gross failure ., f the pressure boundary.

Therefore, the presence of any PRESSURE BMNDARY LEAKAGE requires the _ unit to be promptly placed in COLD SHUTDOWN.

HADDAM NECK B 3/4 4-5 0088

E REACTOR COOLANT SYSTEM BASES Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than I gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.

HADDAM HECK cons B 3/4.4-Sa

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Docket No. 50-213 814087 3

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Attachment 3 i Haddam Neck Plant Significant Hazards Consideration l

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'b July 1992 l

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U.S. Nuclear Regulatory Commission Attachment 3/B14087/Page 1 July 31, 1992 Significant Hazards Consideration Determination .

We have reviewed the proposed change in accordance with 10CFR50.92 and have concluded that the change does involve a significant hazards consideration, g . Specifically, ;he change will:

1 Involve a significant increase in the probability or concm.aences of an accident previon' analyzed. The ieam generator tubes that will be allowed to remain in service will not increase the probability of a s . ear generator tube rupture since their structural

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integrity during normal operattor will not te reduced. The NRC Staff radiological dose assessmeu of the design basis steam line h'+ break analyses considers a primar.y = o secondary leakage of .4 gpm in

< caleplating p.stulated radiological consequences. The proposed

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changes wocid allow steam generator tubes to remain in service such that the postulated primary to secondary leakage following a steam line break could be up to 100 gpm. This would result in a significant increase in the calculated radiological consequences of a steam line break. However, using raore realistic but still h conservative assumptions the postulated radiological cnnsequances 1

wo'.:1d be a small fraction of 10CFR100 limits. Based on this, the prormed changes are considered safe but would,' using NRC Staff dose

~* assumptions, significant1v increase the radiological zes of a steam line break.

the possibility of a new or different kind of accident. The proposed change would allow a postulated primary-to-secondary-leak rate of up to 100 gpm following a steam line break. Currently, the steam line break analyses do not consider an increase in primsry-to-secondar3 leakage following the accident. Allcwing steam generator tubes, that are postulated to leak following a steam line break, to remain in service is considered to introduce the possibility of a malfunction of a different type than previously eva.luated. .

3. Not involve a significent reduction in the margin of safety. The change vill not reduce the margin of safety for Peam generator tube failure. Also, it will not significantly affect the predicted reactor core nor containment _ rerponse following a postulated steam line break.

The safety evaluation concludes that a plant-specific application of generic alternate repair criteria provides the same margin of safety for steam generator tube structural design with-respect to bursts.

The plant-specific application of the gencic alternate plugging criteria provides a calculated theoretical leakage rate for steam generator tubes having cracks that re pere,itted to remain in tervice. The accident which is affected by this proposed change is a Main Steam Line Break. When the calculated theoretical- leakage rate is used to evaluate the dose consequences of a main steam line -

$ U.S. Nuclear Regulatory Commission

- Attachment 3/B14087/Page 2 July 31, 1992 break, dose consecuences are higher than those calculated previously by the NRC -Staff in accordance with Standard Review Plan 15.1.5, Appendix A methodology. This methodology is overly conservative and inconsistent with the original dose calculation methodology associated with the initial licensing of the Haddam Neck Plant. In addition, the results of recent 'research indicates that the SRP 15.1.5, Appendix A methodology may contain overly conservative assumptions. 'CYAPC0 hss revised the radiological assessment methodology acccrdingly. Usirg this revised dose calculation methodology and the calculated theoretical leakage rates, dose contributions from the aggregate tube leakage will be limited to a fraction of 10CFR100 dose guidaline values. Further details are provided in Attachment 1.

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