ML20072Q660

Proposed Tech Specs Allowing Option of Using B&W Kinetic Sleeving Process for 3/4 Inch OD Tube Repair Described in Topical Rept BAW-2045(P)-A

ML20072Q660
Person / Time
Site:	Catawba
Issue date:	12/19/1990
From:	DUKE POWER CO.
To:
Shared Package
ML20072Q650	List: ML20072Q647 ML20072Q660
References
NUDOCS 9012260277
Download: ML20072Q660 (9)

Text

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4 g 6ges NS 98 9d o REACTOR COOLANT SYSTEM d 3/4.4.5 STEAM GENER\ TORS ..

LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one or more steam generators inoperable, restore the inoperable generator (

to OPERABLE status prior to increasing T,yg above 200*F.

SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator j shall be oetermineo OPERABLE during shutdown by selecting and inspecting at least.the minimum number of steam generators specified in Table 4.4-1, 4.4.5.2 Steam Generator Tube Samole Selection and Insoection - The steam generator. tube minimum sample size, inspection result classification, and the e x responding action required shall be as specified in Table 4.4-2, The inservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;

b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:

9012260277 DR 9o1219 l ADOCK 05000413 PDR CATAWBA - UNITS 1 & 2 3/4 4-12

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_ REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

1) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),

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.

c. For Unit 1, in addition to the 3% sample, all tubes for which the alternate plugging criteria has been previously applied shall be inspected in the tubesheet region,

d. 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 tu.be inspection provided:

1) The tubn selected for these samples include the tubes from 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:

l Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none'of the inspected tubes are defective.

l 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 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective, i Note: In all inspections, previously degraded tuoes must exhibit significant'(greater than 10%) further wall penetrations to be' included in the above percentage calculations.

CATAWBA - UNITS 1 & 2 3/4 4-13 Amendment No. 47 (Unit 1)

Amendment No. 40 ' Unit 2)

k No CAam3 es 'Ibi ha3e, REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Conti m d) 4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the

= preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to l a maximum of once per 40 months; l

b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 inonths. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification'4.4.5.3a.; the interval may then be extenced to a , -

maximum of once per 40 months; and o

c. Additional,, unscheduled inservice inspections shall be performed each steam generator in accordance with the first sample inspect 1va specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions: ,

1) Reactor-to-secondary tubes ' leak (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or

2) A seismic occurrence greater than'the Operating Basis Earthquake, or

3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or

4) A main steam line or feedwater line break.

l CATAWBA - UNITS 1 & 2 3/4 4-14

. i REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria a.

As used in this specification: gg e -

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1) Imperfection means contour of a tube k,an romexception to the that required bydimensions, finish or or fabrication drawings specifications. Eddy-current testing indications below 20% of the nominal tube all thickness, if detectable, may be considered as imp fections; o e s\eeJc-

2) Deoradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tubej or s\cco c. w

3) Dearaded Tube means a tube *containing imperfections greater thanorequalto20%ofthenominalgallthicknesscausedby degradation; h ube. ce 5\ceoc

4) -% Oearadation means the percentage of the tube wall thickness -

affected or removed by degradation; lo g- sleco s -

5). Defect means an imperfection of such severity that it exceeds the p1 +ng' limit. A tube containing a defect is defective;

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6) _ hek. Limit means the imperfection depth _at or beyond which --

the tube :. hall be removed from servicFandede equal to 40% of the nominal tubewall thickness. For Unit 1, this definition

• M*"T166s not apply to the region of the tube subject to the alter-nate tube plugging criteria.

V f '7) '. Unserviceable describes the condition of a tube if it leaks or-contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or.a steam line or feedwater line break as specified in 4.4.5.3c., above; 8). Tube Inspection 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; l dred b 5Iuuin M a) s o e n5 M c 3 Iv33+ing or ec .,,,)

W cc. i o c cle d o ,- b c n d w h;ch 2. s(ce_ve.J U.c. Sho.\\ h e . vgech The. VF- 3 o iv'* \'% V iS ->

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~The B a.b c ock 4 Wi lc o /. g rc<e s s de scr b ed i 'l Tof ic" I

]egort Bh to - 2o45 (P)-h 4,11 be u se d for sl e c MS .J CATAWBA - UNITS 1 & 2 3/4 4-15 Amendment No. 47 (Unit 1)

Amendment No. 40 (Unit 2)

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REACTOR COOLANT SYSTEM 1

SVRVEILLANCE REQUIREMENTS (Continued)

9) Preservice Inspection means an inspection of the full length of j each tube in each steam generator performed by addy current bchniques prior to service to establish a baseline condition of G.? tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques  !

expected to be used during subsequent inservice inspections.

10) Tube Roll Expansion is that portion of a tube which has been increased in oiameter by a rolling process such that no crevice exists between the cutside diameter of the tube and the tubesheet.

11) F* Distance is the minimum length of the roll expanded portion of the tube which cannot contain any defects in order to ensure the tube does not pull out of the tubesheet. The F* distance is 1.60 inches and is measured from the bottom of the roll expansion transition or the top of the tubesheet if the bottom of the roll expansion is above the top of the tubesheet.

. Included in this distance is a safety factor of 3 plus,e 0.5 inch eddy current vertical measurement uncertainty.

12) Alternatetubepluandc eria does not require the tube to

. 'beremovedfromsep; ice.orrepairedwhenthetubedegradation exceeds the Trkg;L limit so long as the degradation is in that portion of the tube from F* to the bottom of the '

tubesheet. This definition does not apply'to tubes with s degradation (i.e,, indications of cracking) in the F*

distance, l b.- The steam generator shall be determined OPERABLE.after completing l

-the corresponding actions (plug or repair all tubes exceeding the-R e dP + plugghg-limit and all tubes containing through-wall cracks) required by Table 4.4-2. For Unit 1, tubes with defects below F* fall under the alternate tube plugging criteria and do not have to be plugged.-

4.4.5.5 / Reports gr7e cd.,

H a. Within 15 days following the completion of each in ervke inspection of steam generator tubes, the number of tubes p+ in each steam generator shall be reported to the Commission in a Special Report' p pursuant to Specification 6.9.2; l

L 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 12 montns following the completion of the inspection. This Special Repor_t shall include:

1) Number and extent of tubes-inspected, I

CATAWBA - UNITS 1 & 2 3/4 4-16 Amendment No. 47 (Unit 1)

Amendment No. 40 (Unit 2)

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REACTOR COOLANT SYSTEM _

SURVEILLANCE RE0V!REMENTS (Continued)-

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

3) Identification of-tubes q rc(a +ie d .

c. For Unit 2, 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 within 30 days and prior to resumption of plant op9 ration. This report shall provide a description of investigations conducted to determine cause of the i tube degradation and' corrective measures taken to prevent recurrence. '

d. For Unit 1, the results of inspections for all tubes for which the alternate tube plugging' criteria has been applied shall be reported to-the Nuclear Regulatory Commission in accordance with 10 CFR 50.4, prior to restart of the unit following the inspection. This report- ,

shall include:

1) Ident.ification of applicable tubes, and

2) location and size of the degradation.

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CATAWBA - UNITS 1 & 2 3/4 4-16a Amendment No. 47 (Unit 1)

Amendment No. 40 (Unit '2)

r a-tJw %uy rk Py REACTOR COOLANT SYSTEM ,

BASES .j

' SAFETY VALVES (Continued)

. 4 relief _ capability and will prevent overpressurization..- In addition, the- j Overpressure Protection System provides a diverse means of protection against overpressurizationLat low temperatures.

During -operation, all pressurizer _ Code safety valves must be OPERABLE to ,

. prevent the Reactor Coolant System from being pressurized above-its Safety 1 Limit of'2735 psig. The combined relief capacity of'all of_these valves is greater than the maximum. surge rate resulting from a complete loss-of-load assuming 'no Reactor trip until the first Reactor Trip System Trip Setpoint. is reached (i.e., no credit is taken.for a direct Reactor trip _on the loss-of-load) and also assuming-no. operation of the power-operated relief valves-or steam '

dump valves.

Demonstration _of the safety valves' lift settings -will occur only during -l shutdown and-will: be performed in accordance with the provisions of .Section XI l'

of_ the ASME Boiler and Pressure Code.

3/4;4.3 PRESSURIZER The limit on the maximum water volume in=the pressurizer assures that the parameter is maintained within the normal steady-state envelope-of operation assumed in the_SAR.: The limit..is consistent with' the initial SAR assumptions, i The-12-hour periodic _ surveillance is-sufficient.to ensure that the parameter is restored to'within'its-limit-following expected-transient operation.

The t

maximum' water' volume-also ensurs:s that a steam bubbleLis= formed and thus the- ,

Reactor Coolant -System'is tnot a hydraulically solid sys tem. The requirement that a minimum' number of pressurizer heaters be OPERABLE enhances the capability.

of the plant-to control Reactor Coolant-System pressure and establish natural l circulation.: i

-3/4;4:4" RELIEF ~ VALVES The = power-operated . relief valves =(PORVs) and steam bubble function' to

' relieve Reactor Coolant System pressure =during all. design: transients up to and including the-design step load: decrease with steam dump. . Operation of the-

-PORVs minimizes tre. undesirable opening of the spring loaded pressurizer. Code

= safety valves. Each-PORV has a remotely operated block valve.to_ provide a positive' shutoff-capability should a relief valve become inoperacle. -Testing

, of the PORVs-inclodes the-emergency N: sLpply from the Cold Leg-Accumulators.

This- test demonstrates, that' the valves ir the supply .line . operate. satisfactorily and that the nonsafety portion of the-instrument air system -is not necessary fo-_ properPORV operation.

3/0 4.5 STEAM GEhERATORS The'Surveilitnce Requirements for irspection' of the steam generator tubes ensure that;the structural integrity of tnis portion of the Reactor Coolant System Will be-maintained. The program for inservice inspection of steam CATAWBA - UNITS 1 & 2 8 3/4 4

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i REACTOR'C00LANT SYSTEM BASES STEAM GENERATORS (Continued)

_ generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to main-

-tain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manu-facturing errors, or inservice conditions that lead to corrosion. Inservice inspec' tion of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. '

b _ coolant will be maintained within those chemistry limits found to result inhe

&cWnegligible' corrosion- of the steam generator tubes. If the secondary coolant

- y chemistry is not maintained within these limits, localized corrosion may~likely result in stress corrosion cracking. The extent of cracking during plant opera-tion would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 500 gallons per day per steam generator). Cracks having a reactor-to-secondary leakage less than this limit during operation will have.an adequate margin postulated of safety to withstand the loads imposed during normal operation and by' accidents. .4 Operating plants have demonstrated that reactor-to- I 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  ! ? :;p. res.ir*2 Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a-defect should develop in service, it will be found during scheduled inservice steam generatcr tube examinations.

f M ukQ" u;;;;in L limit of 40% of the tt )e nominal wall' thickness, For Unit 1, irs will be m p Tefecti e: tubes which fall unde

• the alternate tube plu1ging criteria do irefUghave to be Qu;;d.A Steam generator tube-inspections of operating plants have-demonstrated the eftpability to reliably _ detect wastage type. degradation that ha p d 200 of the original tube wall thickness.-

WM[f allWhenever into Category the results of any steam generator.~ tubing inservice inspection C-3, these results will be reported to the suant to' Specification 6.9.2 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 recuirement for analysis, laboratory examinations, tests, additional L eddy-current inspection necessary. I.n se+ q, , and _ revision of the Technical Specificat ions, if 3/4.4.6' REACTCR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 ~LEAKAGI DETECTION SYSTEMS The Leakage letectic Systems recuir ed by this specificatior are pre.ided-ta monitor and detect lean age from the rt actor co:lant pressure t oundary.

CATAWBA - UNITS 1 & 2- B 3/4 4-3 A nendment Nc. 47 (Unit 1)

Amendment No. 40 (Unit 2) l

Attachment No. 2 hases, 3/4.4.5 St eam Generators, insert item A The B&W process (or method equivalent) to the inspection method described in Topical Report BAW-2045(P)-A will be used. Inservice inspection of steam generator sleeves is also required to ensure RCS integrity. Because the sleeves introduce changes in the wall thickness and. diameter, they reduce the sensitivity of oddy current testing, therefore, special inspection methods must be used. A method is described in Topical Report BAW-2045(P)-A with supporting validation data that demonstrates the inspectability of the sleeve and underlying tube. As required by NRC for licenscos authorized to use this repair-process, Catawba com its to validate the adequacy of any system that is used for periodic inservice inspections of the sleeves, and will evaluate and, as deemed appropriate by Duke Power Company, implement testing methods as better methods are developed and validated for commercial use.

Rasos, 3/4.4.5, Steam Generators. Insert item B:

Defective steam generator tubes can be repaired by the installation of-sleeves which span the area of degradation, and serve:as a replacement prosst"o boundary for the negraded portion of the tube, allowing Lho tube to remain in service.

Bases, 3/4.4.5. Steam Generators, Insert Item C:

If a. tube is s1 coved duo to degradation in the F* c1 stance, then any defects in the tube'below the sleeve will remain in servico without repair.

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