ML20128E550

Proposed Tech Specs Re Steam Generator Repair Criteria

ML20128E550
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	01/29/1993
From:	CONNECTICUT YANKEE ATOMIC POWER CO.
To:
Shared Package
ML20128E546	List: B14327, Application for Amend to License DPR-61,revising TS Re Steam Generator Repair Criteria,To Reduce Unnecessary Plugging of SG Tubes in Tubesheet Expansion Zone Roll Transition Area ML20128E550
References
NUDOCS 9302100510
Download: ML20128E550 (11)

Text

_-_-____ __ ___

c Docket No. 50-211

! B14327 Attachment 2 Haddam Neck Plant Propossa Revision to Technical Specifications January 1993 9302100510-930129

.PDR ADOCK 05000213 P PDH

REACTOR COOLA!!T SYSTEM SORVEILLANCE RE0VIREMENTS (Continued)

1) All nonplugged degraded tubes. -

l

2) Tubes in those 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 inspection, 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, i 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:

L 1) The tubes selected for these samples include the tubes from l- those areas of 'the tube sheet array where tubes with imperfections were previously found, and

2) The inspections include those portions of the tubes where-imperfections were previously found.

The results of each sample inspection shall be classified into one 'of the

• following three categories:

Cateaory Inspection Resulti C-1 Less than - 5% of. the totai tubes inspected e.re degraded 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.

l 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 inspected 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- percen)tagel calculations.This does not appiy within the tube-to-tubesheet' roll region.

HADDAM NECK.

cons 3/4 4-23 Amendment No. JU

REACTOR COOLANT SYSTEM L

i- S M EILLANCE RE0VIREM W i (Continued) _

4.4.5.4 Acceptance Criter.ia

a. As used in this specification:

1) Imoerfection means an exception to the dimensions, finish of contout of a tube from that required by fabrication drawings or specifications. Eddy-current- testing indications below 20% of- the nominal tube wall thickness, if detectable, may be considered as imperfections;

2) Dagradation 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 l free of imper fections; i

4) Dearaded Tube means a tube containing imperfections 1 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 I 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)  % Deoradatign means the percentage of the tube wall ' l thickness affected or removed by degradation;

6) Defect means an imperfection of such severity :that it exceeds the slugging limit.- A tube or sleeve containing a defect is cefective;

7) Pluacina Limit means the imperfection depth at or beyond l-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 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 roll 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:

HADDAM NECK 3/4 4-25 AmendmentNo.J#

0087

T REACTOR COOLANT SYSTEM SEVEILLANCE REOUIREMENTS (Continued) ._

1) 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 ~

circumferentially 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 100FR100 dose guideline values in the event of a-steam line break.

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

s

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IIADDAM NECK- 3/44-25a- Amendment Na 6087

REACTOR COOLANT SYSTEM jil!RVEILLANCE RE0VIREMENTS (Continued)

7) Unserviceable describes the condition of a tube if it contains {

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) Iube insoection means an inspection of . the rtea'm generator tube from the hot les entry point completdy 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 around the U-bend to the oppcsite end,

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

• all tubes exceedtng 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 ,Recorts

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;

b. The complete results of the steam generator tube inservice inspection shall be submitted to -the Commission in a Special Report pursuant to Specification 6.9.2 within 90 days following the completion of the inspection. This Special Report shall include:

i) Number and extent of tubes inspected,

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

3) Identification of tubes plugged, sleeved, or dispositioned i 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 Cor.aission 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.

• lube sleeving shall be performed in accordance with the connecticut Yankee Steam Generator Sleeving Report, Revision 1, transmitted by letter dated January 7,1986.

1 HADDAM NECK 3/4 4-26 Amendment No. JEE oasi

l REAQJDR COOLANT SYSTEM' I BASES 3/4.4.5 STEAM GENERATORS (Continued) l over a large araa. 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 i problem. Of course, these tubes are inspected each outage to measure. their i defect size, and they are plugged or repaired if necessary.

1 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 stresses which can lead to stress corrosion cracking of the tube. Those tubes which do not permit the passage of a 0.460 inch diameter probe are plugged.

Stress corrosion cracking is caused by a combination of stress in the tube with or without coincident adverse cher. .stry. Once a stress corrosion crack begins to form, it may propagate rapioly. 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 leakage 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 gallot per day per steam generator can readily be detected by radhtion 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 defined in Specification 4.4.5.4.a.7. I 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 detacted during inservice inspections, and are plugged or- repaired if- the defect size exceeds 50%.

Al so, leakage caused by pitting during operation would be monitored and measured as discussed above. Testing has showr that pits will cause tube leaks before affecting tube integrity.

If a defect should develop in service, it would be found during scheduled inservice steam geaerator tube examinations.- Plugging or sleeving will be required for all tubes with imperfections equal to or exceeding the plugging.

limit as defined in Specification 4.4.5.4.a 7. Tubes containing sleeves with im3erfections -exceeding the plugging limit will be plugged. Steam generator tu)e' inspections of operating plants have demonstrated the capability - to reliably detect degradation 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.

4 HADDAM NECK - B 3/4 4-4 Amendment No. 11%

0088

' REACTOR COOLANT SYSTEM EASES

-3/4.4.5 STEAM GENERATj),81_Kontinued)

The alugging or sleeving limit defined in Specification 4.4.5.4.a.7 is. based I on tie requirements set forth in Regulatory Ge de 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 sound roll or is characterized as a crack in the roll expansion region which meets. the requirements defined in Specification 4.4.5.4.a.7.b. The number of 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 of 10CFR Part 100 dose guideline values in the event of a steam line break. Using conservative assumptions, a postulated post accident leak rate of 1.26 gpm per steam generator would result in doses 10% of 10CFR100 limits.

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

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

into Category 'C-3, these results will be reported to the Commission in a 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 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.

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 I gallon per minute in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as ' stated in Condition (2) of Generic Letter 84-04. Because the 51ume Control Tank Level Monitoring System and the Containment 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 specificatina are provid-ed to monitor and detect leakage from the reactor coolant.pr' re 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 LEAKAGI PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since -it may be indicative of an impending gross fa! hwa of the pressure boundary.

Therefore, the pruence of any PRESSURE BOUNDARY LEAKAGE requires the ' unit to be promptly placed in COLD SHUTD0VN.

HADDAM NECK B 3/4 4 Amendment No. W 0088 m

a REACTOR CQDLANT SYSTEM EASES 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 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.

HADDAM NECK B 3/4 4-5a Amendment No.

0088

p H

. . . . - _.1 Docket'No. 50-213 B14321

.o Attachment 3 Haddam Neck Plant Significant Hazards Consideration January 1993

. 1.m - 1

U.S. Wuclear Regulatory Commission Attachment 3/B14327/Page 1 January 29, 1993 Sionificant Hgards Consideration DeterminaMqn We have reviewed the proposed change in accordance with 10CFR50.92 and have concluded that the change does not involve a significant hazards consideration.

Specifically, the change will not:

1. Involve a significant increase in the probaoility or consequences of an accident previously analyzed.

The steam generator tubes that will be allowed to remain in cervice ,

will not increase the probability of a steam generator tube rupture since their structural integrity during normal cperation and postaccident conditions as required by Regulatory Guide 1.121 will not be reduced. The :lRC Staff radiological dose assessment of the design basis steam line break analyses considers a primary to secondary leakage of .4 gpm in calculating postulated radiological consequences. The proposed changes would allow steam generator tubes to remain in service such that the postulated primary to I secondary leakage following a steam line break could be up to 1.26 gpm per steam generator. This would result in an increase in the calculated radiological consequences nf a steam line break, cut still would be a small fraction of 10CFR100 limits.

The alternste repair criteria being applied at Haddam Neck is considered conservative since this methodology for the revised plugging criteria was developed by a committee of industry experts under the cognizance of EPRI wh;ch considered the effects of crack size and location of the cracks to ensure that the tubes remain structurally sound and that leakage that may occur remains well within the required values. The Haddam Neck design is such that it offers more protection than that contained in the EPRI documents.

This is as a result of the design of the steam generators at Haddam Neck. The generic alternate plugging criteria assumed that the steam generator tubes are fully expanded through the tubesheet and-that the cracks are located above the top o' the tubesheet. The generic criteria has length limitations to preclude burst of the cracked tube. Application of this crl+.eria at Haddam Neck results in an extremely conservative result since the steam generator tubes are only partially expanded through the tubesheet. The tubesheet completely surrounds the tube in the' region where the cracks are located, precluding the bursting of the tubes.

Failure modes other than bursting of the tubes were also investigated. The complete severance of the tube at the crack location and subsequent pullout of the tube from the tube cheet was evaluated. The tubes were deemed to remain structurally sound under normal and accident conditions. Partial collapse during a less of coolant accident was also investigated and was able to be excluded

s e . .,

U.S. Nuclear Regulatory Commission

-Attachment 3/B14327/Page 2 January 29, 1993-based on analysis and testing. Based on this, the proposed changes do not significantly increase the probability or consequences of an accident previously evaluated.

2. Create 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 1.26 gpm per steam generator following a' steam line break. Currently, the steam line break analyses consider a primary-to-secondary leakage design basis of 0.4 gpm following the accident, but does not consider the potential for an increase in leakage. A leak rate of 1.26 gpm will not modify plant or equipment response to an SLB, The potential for cracks or breaks in the steam generator tubes have been studied by both EPRI and CYAPCO. CYAPC0 has concluded that the leakage through the allowable cracks will maintain the plant within the safety accep'.ance criteria. In addition, bursting, pulling out or collapsing of tubes has been ,

analyzed and will not occur. Therefore, this postulated type of-leak and leak rate do not create- the possibility of a new or different type of accident.

3. Not involve a significant reduction in the margin of safety.

The change will not reduce the margin of safety for staam generator tube failure. Also, it will not affect the predicted reactor core nor containment response 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 generic alternate repair criteria provides a calculated theoretical leakage rate for steam generator tubes having cracks that are permitted to remain in service. The accident which is affected by this proposed change is a Main Steam Line Break. However, the calculated off-sito dose consequences based on the theoretical-leakage linit (1.26 gpm per steam generator) is significantly less than 10CFR100 limits.

In summary for the reasons identified above, CYAPC0 has concluded that continued-operation of the facility in accordance with the proposed amendment would not involve a significant hazards consideration.

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