ML20096G998

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Proposed Tech Specs 3.4.5,modifying SG Tube Plugging Criteria to Allow Implementation of Alternate SG Tube Plugging Criteria for tube-to-tubesheet Joints (Known as F*)
ML20096G998
Person / Time
Site: South Texas STP Nuclear Operating Company icon.png
Issue date: 01/22/1996
From:
HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML20096G996 List:
References
NUDOCS 9601260218
Download: ML20096G998 (7)
v • d • e


Text

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ATTACHMENT 3 MARKED UP PAGES FOR PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS j

TSC-093270 9601260218 960122 PDR ADOCK 05000498 P PDR

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ST-HL-AE-5270 Attachment 3 Page 1 of 5 REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) *

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.
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 i provided: 1 l
1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously l

found, and l

2) The inspections include those portions of the tubes where imperfections were previously found, d[y"!F6EUniC17ssi'lubi*sil6EditsTremaisis'seWi6iipd^Xcceptance1Criterioillo

{ofsTshuicsl:Specificati6n;;44;5:4)lshall bel inspected lvia;theirotating pnM l Tsddysurrent theth6d'byst th(ENdistance.)Such; tubes;'are esempt c61!;(RPC)surreht frodieddi isspectioisveEthhip~6rtion*ef:theltube:belonthe

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

Catecorv Insocction Results C-1 Less than 5% of the tote tubes inspected are 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 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 percentage calculations.

SOUTH TEXAS - UNIT 1 & 2 3/4 4-13 TSC46\5270

1 ST-HL-AE-5270 a Attachment 3 Page 2 of 5 REACTOR COOLANT SYSTEM - l i

STEAM GENERATORS l SURVEILLANCE REQUIREMENTS (Continued) q I

i 4.4.5.4 Accentance Criteria  ;

a. As used in this specification:
1) Imoerfection means an exception to the dimensions, finish, or contour of - -l a tube from that required by fabrication drawings or specifications.  ;

Eddy-current testing indications below 20% of the nominal tube wall j thickness, if detectable, may be considered as imperfections;

2) Degradation means a service-induced cracking, wastage, wear, or  :

general corrosion occurring on either inside or outside of a tube; 5 l

3) Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation;  ;

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4)  % Degradation means the percentage of the tube wall thickness affected or removed by degradation; {
5) Defect means an imperfection of such severity that it exceeds the l plugging limit. A tube containing a defect is defective; l
6) Plunnine Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal  !

tube wall thickness, ,

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7) - Unserviceable describes the condition of a tube ifit leaks or 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) Tube Insoection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and
9) Preservice Inspection means an inspection of the full length of each tube  ;

in each steam generator performed by eddy current techniques prior to '

service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.

SOUTH TEXAS - UNIT 1 & 2 3/4 4-15

' TSC.96\S270

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ST-HL-AE-5270 l Attachment 3  !

l Page 3 of 5 l 1

i L REACTOR COOLANT SYSTEM STEAM GENERATORS  !

SURVEILLANCE REQUIREMENTS (Continued) e

~fgJF'GiMHiiTF6FUsif!MisisTulE7deigiudsEn 10 belowisisposifkid i' "diat= ace from the'.hardtroll: contact puint"st Weissittsbof-tubesliest (tts$fdisamaATsantbsisas:Indsd;frus?canaidarntida!.tp t enteria. stated!in'this'sectionl(igjdsgging[6fsuchitubes'is?no(

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b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tube exceeding the plugging- l limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Reoorts l 1

a. Within 15 days following the completion of each inservice  ;

inspection of steam generator tubes, the number of tubes plugged in each steam 4 generator shall be reported to the Commission in a Special Report pursuant to .;

Specification 6.9.2;

b. The complete results of the steam generator tube inservice inspection shall be submitted to the Conunission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special 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.

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c. Results of steam generator tube inspections which fall into Category C-3 shall i be reported in a Special Report to the Commission pursuant to Specification I 6.9.2 within 30 days and 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.

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l SOUTH TEXAS - UNIT 1 & 2 3/4 4-16 neem l

!: ST-HL-AE-5270 Attachment 3 Page 4 of 5 j REACTOR COOLANT SYSTEM -l BASES  !

RELIEF VALVES (Continued)  ;

l i C. Manual control of the block valve to: (1) unblock an isolated PORV to allow it to be-  !

used for manual control of reactor coolant system pressure (Item A), and (2) isolate the PORV with excessive seat leakage (Item B). j 5

D. Manual control allows a block valve to isolate a stuck-open PORV.  !

3/4.4.5 STEAM GENERATORS f The Surveillance Requirements for inspection of the Steam generator tubes ensure that l 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 i 1.83, Revision 1. 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 i provides a means of characterizing the nature and cause of any tube degradation so that l corrective measures can be taken. >

. The plant is expected to be operated in a manner such that the secondary coolant will

.be maintained within those chemistry limits found to result in negligible corrosion of the

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steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (primary-to-

- secondary leakage = 500 gallons per day per steam generator). 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 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

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i Unit 1 - Amendment No. 55 SOUTH TEXAS - UNITS 1 & 2 - B 3/4 4-2a Unit 2 - Amendment No. 44 TSC46\$270

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ST-HL-AE-5270 I

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Attachment 3 i Page 5 of 5 j REACTOR COOLANT SYSTEM l BASES l l  :

j STEAM GENERATORS (Continued) i Wastage-type. defects are unlikely with proper chemistry treatment of the secondary - l coolant. However, even if a defect should develop in service, it will be found during j scheduled inservice steam generator tube examinations.- Plugging will be required for all .  ;

tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall  !

thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall i' thickness.

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Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly 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 Commission 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. j 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE j 314.4.6.1 LEAKAGE DETECTION SYSTEMS i

The RCS Leakage Detection Systems required by this specification are provided to l monitor and detect leakage from the reactor coolant pressure boundary. These Detection l Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

, 3/4.4.6.2 OPERATIONAL LEAKAGE l

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, i

SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-3  !

f TSCA6\5270 l ,

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ATTACHMENT 5 l W E-Series F* Qualification Report BAW-10203 I

TSC 06\5270

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