ML20196F977

Proposed Amend to Facility OLs DPR-39 & DPR-48 to Allow Steam Generator Tube Sleeving

ML20196F977
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	03/02/1988
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML20151M120	List: ML20151M115 ML20151M119 ML20151M128 ML20196F977
References
CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-88-3002, NUDOCS 8803040290
Download: ML20196F977 (25)

Text

- .. / fo c I~ w .o n ' 2 at SAIC-88/3002 TECHNICAL EVALVATION REPORT

, ZION NUCLEAR POWER STATION UNITS 1 AND 2 l

, PROPOSED AMENDMENT TO FACILITY OPERATING  !

! . LICENSE N0s. DPR-39 AND DPR-48 TO ALLOW STEAM GENERATOR TUBE SLEEVING i l

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2 March 1988 )

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t Prepared for:

U.S. Nuclear Regulatory Commission

- Washington, D.C. 20555 Contract NRC-03-87-029 u Task Order No. 09 9 . 0

TABLE OF CONTENTS Section Elat

1. Introduction ........................................... 1 ,

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2. Background ............................................. 1 2.1 Zion Unit 1 ....................................... 2

, 2.2 Zion Unit 2 ....................................... 2

3. Discussion ............................................. 3 l

i 3.1 Sleeve Design ..................................... 5 3.2 Sleeve Material ................................... 6 3.3 Material Corrosion Testing ........................ 7 3.4 Upper and Lower Joints Corrosion Testing .......... 8 3.5 Qualification Testing and Analysis ................ 10 j 3.5.1 Leak Verification Tests ................. 10 3.5.2 Analytical Verification . . . . . . . . . . . . . . . . . 13 u

3.5.2.1 Structural Analysi s . . . . . . . . . . . 13 3.5.2.2 Tube Vibration Analyses ....... 15

, 3.5.2.3 Effect of Flow Slot Hour-glassing Analyses ............. 15 l

3.5.2.4 Allowable Sleeve Degradation Analyses ...................... 16 3.5.2.5 Plugging Limit Determination .. 17 3.6 Eddy Current Testing .............................. 18 i 3.7 Inservice Inspection Plan for Sleeved Tubes ....... 20

4. Evaluation ............................................. 20

5. C o n cl u s i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 References ............................................. 23 L

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1. INTRODUCTION i

On December 24, 1987 Comonwealth Edison Company submitted a proposed amendment to Zion Nuclear Power Station Units 1 and 2 Facility Operating License Nos. DPR 39 and DPR 48, Appendix A. Section 4.3, Reactor Coolant System. This change would allow using a Westinghouse Electric Corporation I mechanical sleeving methodology for Zion's steam generator tube repair, if

, needed. If this amendment request is approved, Comonwealth Edison Company would be able to use either the sleeving methodology developed by Westinghouse Electric Corporation or that developed by Combustion Engineering, Inc, which was previously approved by the Nuclear Regulatory Comission staff.

This report provides a technical evaluation of the licensee amendment request and is based primarily upon the Westinghouse report WCAP 11669,

' Zion Units 1 and 2, Steam Generator Sleeving Report (Mechanical Sleeves),"

which is attachment 4 of the December 24, 1987 request package. Additional resource material is listed in the reference section of this report.

2. BACKGROUND Zion Units 1 and 2 use pressurized water reactor (PWR) steam generators that are manufactured by Westinghouse Electric Corporation. Both units have four Westinghouse Series 51 vertical recirculating U bend tube steam generators, and each has 3388 Inconel 600 tubes. The tubes have a 0.050 inch nominal wall thickness with an outside diameter of 0.875 inches. Zion 1 initiated comercial service in December 1973 and Zion 2 in September 1974.

Both Zion units have experienced steam generator tube degradation in the tubesheet area. This degradation it believed to be corrosion induced and caused by intergranular attack (MA). Because Zion Unit I has experienced one of the highest rates of tube plugging in operating

- Westinghouse Model 51 steam generators, Comonwealth Edison Company sutaitted a proposed Technical Specification change to the NRC on April 24, 1986 that would allow Combustion Engineering welded sleeves to be installed as an alternative to plugging for steam generator repair. The sleeves were l to be installed in tubes degraded near the tubesheet. On November 18, 1986 i

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<O the NRC approved Comonwealth Edison Company's request to perform tube sleeving according to the Combustion Engineering procedure.

2.1 71on Unit 1 The 1986 steam generator tube inspections at Zion Unit I were conducted in September and October 1986. At this inspection 100% of the tubes were eddy current tested using multifrequency examination techniques. The results of these inspections were as follows.

Steam Generator IA 18 1C 10 Tubes Tested 3253 3172 3241 3244 f Defective Tubes 7 31 15 38 "Squirrels" 4 68 9 4 i 'DRIs" 8 1 26 0

! "Squirrel" indications are large magnitude unidentified indications believed to be IGA. Distorted Roll Indications ("DRIs') are believed to be g

possible indications of distorted roll transitions. All defective tubes, t ' squirrels" and "DRIs" were plugged or sleeved.

! At the end of 1986 there were 738 plugged tubes (5.7%) and 128 sleeved tubes at Zion Unit 1. Zion t' nit 1 is scheduled for a refueling outage beginning on February 25, 1988 at which time steam generator tube inspections will be performed again.

2.2 71on Unit 2 Steam generator tube inspect > s were also conducted at Zion Unit 2 l during a refueling outage between September 5,1985 and January 22, 1986.

During these inspections, Comonwealth Edison Company performed multifrequency eddy current examinations of all four steam generators. All the tubes in each steam generator were examined from the hot leg side through the U bend and approximately 900 tubes in each steam generator were examined full-length (eitty from the hot leg exiting through the cold leg).

As a result of these examinations 3 tubes in steam generator 2B and 1 tube in steam generator 2C were plugged. At the end of 1986 there were 40 l 2

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.O plugged tubes and no sleeved tubes in the Zion Unit 2 steam generators.

3. DISCUSSION 1

Historically, when steam generator tube wall degradation is severe, the corrective action taken is a repair technique called plugging. Plugs are installed at the inlet and outlet of the steam generator tube when the reduction in wall thickness reaches a calculated value, referred to as the  ;

o plugging limit. Eddy current testing (ECT) has been used to measure steam l generator tube degradation. The tube plugging criteria include a margin for  !

ECT measurement uncertainty.

The installation of steam generator tube plugs removes the heat transfer surface of the plugged tube and the primary coolant flow rate I available for core cooling is reduced. An alternative to plugging is sleeving. The installation of steam generator tube sleeves does not significantly affect the heat transfer capability of the tube being sleeved.

Furthermore, a large number of sleeves can be installed without  ;

significantly affecting primary flow rate, l

l The decision to use the sleeving technique can be made if two objectives are met. The sleeve must maintain the structural integrity of l the steam generator tube during normal operating and postulated accident i conditions. Additionally, the sleeve must prevent leakage in the event of a throughwall flaw of the original steam generator tube. Tests and analysis must be performed to demonrtrate the capability of the sleeves to perform these functions under normal operating and postulated accident conditions.

't Plugs still have to be installed in degraded steam generator tubes when the tubes cannot be successfully repaired with sleeves. Since tube plugging

, results in a reduction of unit power, and if plugging becomes excessive may require steam generator replacement to restore plant reliability and l l availability, the sleeving process is receiving close scrutiny to determine

} acceptability of various methodologies.

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,. By 1986, Commonwealth Edison Company had plugged over 700 tubes in the

Zion Unit I steam generators due to environmental degradation near the i tubesheet areas of the tube bundle. At that time the utility requested and 3 ,

.s was granted a Technical Specification change that allowed repair of degraded steam generator tubes in both Zion units through the installation of Combustion ingineering welded sleeves.

One of the reasons for the amendment request given by Comonwealth ,

Edison Company in its December 24, 1987 letter to the NRC was that if either the sleeving methodology developed by Westinghouse or that approved for Combustion Engineering sleeves could be used, the utility would have the flexibility of producing a more competitive atmosphere surrounding the procurement of sleeving services for the Zion station.

The Westinghouse Electric (WE) mechanical sleeving process described in WCAP 11669 has been previously used in operating WE steam generators (Indian Point 3, Point Beach 2, and Millstone Unit 2 - Attachment 2 of Reference 5),

and can be sumarized as follows.

The WE steam generator sleeving program for Zion 1 and 2 would involve l the installation of thermally treated Alloy 690 sleeves in both hot and cold legs of the steam generators. The tubes to be sleeved are those where tube

, degradation in the tubesheet area and just above the top of the tubesheet l has exceeded the Technical Specification plugging limit of 40% throughwall indications. The sleeve projects from the bottom of the tubesheet to a f point above the secondary side of the tubesheet. Two sleeve lengths would be used in the sleeving process. Both are long enough to span the degraded

, areas of the tubing in the tubesheet region in either the hot or cold legs.

The sleeve is to be secured in the tube by mechanical joints at the top and the bottom of the sleeve.

The WCAP 11669 report describes laboratory testing and actual field performance used to qualify the sleeve design and installation process.

! i This report also describes analytical verifications that have been performed using design and operating transient parameters to envelop loads imposed l

during noma 1, operating, upset and accident conditions. Fatigue and stress I analysis of sleeved tube assemblies in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section III are also described.

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d Non proprietary details of the WE sleeve design, mechanical sleeving process and results of qualification testing, analyses and plant operating experience are sumarized below.

The WCAP 11669 report states that the Zion steam generators were built to the 1965 Edition of Section III of the ASME Boiler and Pressure Vessel Code; however, the sleeves have been designed and analyzed based upon the  :

1983 Edition of Section III of the Code through the Winter 1983 Addenda as well as applicable Regulatory Guides. The associated materials and  ;

processes also meet the Code requirements and criteria. The following table is taken directly from the report.

ASME Code and Regulatory Requirements ,

item Aeolicable Criteria Reouirement Sleeve Design Section III NB-3200, Analysis NB-3300, Wall Thickness Reg. Guide 1.83 S/GTubeInspectability l

I Reg. Guide 1.121 Plugging Margin

}' Sleeve Material Section II Material Composition Section III NB-200, Identification, Tests &

Examinations Code Case N 20 Mechanical Properties Sletve Joint 10 CFR 100 Plant Total Primary / Secondary l Leak Rate l Technical Specifications Plant Leak Rate

3.1 Sleeve Desian Details of the WE sleeve design are proprietary and therefore only the non proprietary design descriptions are sumarized in this report.

At the upper end of the sleeve a joint described as a hybrid expansion l joint (HEJ) is produced by dual expansion techniques. The joint at the  !

lower end of the sleeve is also developed by multi-expansion processes. The 5

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lower end of the sleeve has a preformed section to facilitate the seal formation and to reduce residual stresses in the sleeve.

The following description of sleeve design features was provided by Westinghouse:

"The sleeve, after installation, extends above the top of the tubesheet and spans the degraded region of the original tube. The sleeve length is controlled by the insertion clearance between the channel head inside surface and the primary side of the tubesheet, and the tube l degradation location above the tubesheet. The remaining design h

parameters, such as wall thickness and material, are based on design margins, corrosion resistance, and ASME Boiler and Pressure Vessel Code requirements.

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, "The upper joint is located to provide a length of free sleeve abcve it. The length is added so that if the existing tube were to become J

severed just above the upper edge of the mechanical joint within the I free sleeve region, the tube would be restrained by the sleeve from lateral motion, and therefore, axial motion and subsequent leakage l would be limited.

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! "The upper end of the sleeve is tapered in the thickness to reduce the effect of the double wall in the eddy current signal interpretation.

j , ... To minimize stress concentrations and enhance inspectability in the areas of the upper expanded region..." other sleeve design factors are imposed. I i

i 3.2 Sleeve Material The raterial used for the original generator tubes and the proposed I sleeve material are described in WCAP 11669:

l "The material of construction of the steam generator tubes of the Westinghouse design, including the Zion steam generators, is Alloy 600 1

in the mill annealed (MA) condition. Inconel 600 is a high nickel austenitic alloy that is nominally 725 nivel,14177, chromium, and 6-107, iron. The sleeve material proposeo for sleeving the Zion steam l

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generators is Alloy 690 in the thermal treated (TT) condition. Inconel 690 is also a high nickel austenitic material but contains a higher chromium content and a correspondingly lower nickel content than the l

Inconel 600. It has a nominal composition of 60% nickel, 30% chromium  !

and 9% iron."

3.3 Material Corrosion Testina The Westinghouse report compares and evaluates the corrosion properties i of thermally treated Inconel 600 and 690 (Alloys 600 TT and 690 TT) and mill annealed Inconel 600 (Alloy 600 MA).

1 "Alloy 690 TT is recomended in lieu of Alloy 600 MA or Alloy 600 TT.

Laboratory testing has shown that Alloy 690 has resistance to corrosion in steam generator environments that is equal to or better then Alloy

600 in either the mill annealed or thermal treated ccndition. The higher chromium content of Alloy 690 is believed to be responsible for this enhanced corrosion resistance. In addition, the alloy is thermally treated to further enhance its stress corrosion cracking (SCC) resistance properties.

t "The SCC performance of thermally treated Alloys 600 and 690 in both off chemistry secondary side and primary side environments has been extensively investigated. Results have demonstrated the additional stress corrosion cracking resistance of thermally treated Alloys 600 ,

and 690 compared to mill annealed Alloy 600 naterial. Direct comparison of thermally treated Alloys 600 and 690 has further indicated an additional margin of SCC resistance for thermally treated i Alloy 690." A sumary of corrosion cemparison data for thermally '

treated Alloys 600 and 690, based on test results reported by the licensee, is as follows:

' Thermally treated Alloy 600 tubing exhibits additional SCC and IGA resistance in both secondary side and primary side environments when compared to the mill annealed condition.

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"Thermally treated Alloy 690 tubing exhibits additional SCC resistance compared to thermal treated Alloy 600 in caustic, acid sulfate, and primary water environments. )

l "The alloy composition of Alloy 690 along with a thermal treatment provides additional resistance to caustic induced IGA. l i

"The addition of 10% Cu0 to a 10% deaerated NaOH environment

. reduces the SCC resistance of both thermally treated Alloy 600 and 690. Lower concentrations of either Cu0 or NaOH had no effect, nor did additions of Fe3 40 and SiO2." The performance of Alloy 690 was always equal to or better than Alloy 600.

"Thermally-treated Alloy 690 is less susceptible to sensitization ,

than Alloy 600." l "Continuing investigation of the SCC resistance of Alloys 600 and 690 j in primary water environments has shown mill annealed Alloy 600 to be j susceptible to cracking at high levels of strain and/or stress. j

, Thermal treatment of Alloy 600 in the carbide precipitation region l enhances its SCC resistance. The performance tests of Alloy 690, both j mill annealed and thermally treated, demonstrate primary water SCC l resistance that results from alloy composition.'

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'he WCAP 11669 report states, "All the points summarized above that are l relat eve to the corrosion and SCC resistance of of thermally treated Alloys 600 and 690 are considered applicable to the sleeve."

3.4 Vooer and lower Joints Corrosion Testina  !

Westinghouse conducted a corrosion verification test program, similar to the materials corrosion tests, to demonstrate that the residual stresses induced in the parent tubing by the expansion process do not degrade the integrity of the tubing.

6 The expansion processes for both the upper and lower joints involve a j combinat'on of expansions designed by Westinghouse to minimize tensile I 8

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. f 'l stresses in the sleeve and the outer tube and hence reduce crecking in primary water and corrosive media.

Testing for cracking in standardized aggressive corrosive media is one method of establishing the presence of high tensile stresses in alloys such as austenitic stainless steels and Inconels.

The WCAP 11669 report states: "Confirmation that the outer diameter (OD) stresses on the parent tubing are very low tensile or compressive w s obtained by X-ray diffraction analysis of en Inconel 600 tube expanded 30 mils and by the. parting / layer remov'1 technique." Stress corrosion cracking tests in standardized corrosive m76 4 also established thht the residual stresses on the outer diammr of the tube are compressive.

Westinghouse states: "The residual stresses in a HEJ with an Inconel 600 MA tube /Inconel 690 TT sleeve were measured using the narting/ layer removal technique. The OD surface of the tube was in compression in the ,

axial direction at all locations along the expansion transitions and was f

, i compressive on the ID surface of the +ube except for one measurement in the j unexpanded tube near the transition that was in tension (about 5 ksi). The i i OD surface of 'he sleeve was also in compression in both the axial and circumferential directions except for one measurement that was in tension (about 5 ksi) in..." one expanded region. "The ID of the sleeve had areas where the stresses were as high as about 25 ksi in either the axial or I circumferential direction. Residual stresses of this magnitude should r.ot affect the service performance of the special thermally treated sleeve material."

The results sumarized above showed good correlation with SCC tests and X-ray measurements.

WCAP 11669 presents a series of corrosion tests and test results involving 1) corrosion and stress corrosion cracking of upper joints, 2) corrosion and stress corrosion cracking of lower sleeve joints and 3) corrosion and stress corrosion cracking in the annulus.

Using Type 304 stainless steel and Inconel 600 in the sensitized, mill annealed and thermally treated condition, production type sleeve / tube 9 .

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o configurations were corrosion tested in primary water, and in acid and caustic corrosive medie..

The proprietary test details and test results were reviewed, and it appears that the Westinghouse sleeving process does not impose the high residual tensile stresses that would promote stress corrosion cracking in susceptible tube or sleeve material in any of the test media.

3.5 Oualification Testino and Analysis Westinghouse states: "Qualification testing was conducted to (a) verify the leak resistance of the upper and lower sleeve-to-tube joints, (b) verify the structural strength of the sleeve under normal and accident conditions, (c) verify the fatigue strength of the sleeved tube undar transient loads representing the remaining design life of the plant, ... (d) establish the parameters required to achieve satisfactory installation and performance..." and (e) confirm capability for installation of sleeves in l

tubes with conditions such as deep secondary side hard sludge and tubesheet 6 denting.

! Westinghouse further states: "The acceptance criteria used to evaluate the sleeve performance are leak rates based on the plant Technical Specifications (TS). Testing encompassed static and cyclic pressures, temperatures, and loads. The testing included evaluation of joints fabricated using Inconel 600 sleeves, and Inconel 690 sleeves in the Inconel 600 tubes. While the bulk of the original qualification data is centered on Inconel 600 sleeves, a series of verification tests were run using Inconel 690 sleeves to demonstrate the effectiveness of the joint formation process and design with either material."

l 3.5.1 Leak Verification Tests Westinghouse conducted extensive testing on both upper and lower sleeve-to-tube expansion joint specimens including a mock-up test that was representative of the partially rolled tube-to-tubesheet joint in the Model 10

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s0 44/51 steam generators. Westinghouse states that the Zion steam generator tubes are:

... full depth rolled explosively expanded inside the tubesheet. The f formation of the lower mechanical rolled joint of the tube / sleeve is expected to be identical for both partially rolled tubes and full depth rolled tubes. Then the preformed sleeve (thermally treated Inconel 600 or thermally treated Inconel 690) was inserted into the tube and the lower joint formed.

"The verification test program for the upper joint was similar to that for the lower joint. The upper joint was subjected to fatigue loading cycles and temperature cycles to simulate five years of normal operation and the leak rate was determined before and after this simulated normal cperation."

t I Site specific or bounding analyses were performed by Westinghouse to p

determine the allowable leakage per sleeve durhg normal operation and the I limiting postulated accident condition. Westinghouse states: "The leak rate criteria are based on Technical Specification and Regulatory

! requirements. The leak rate for normal operation is based on the 500 gallons per day (0.35 gpm) per steam generator limit in the Zion Unit I and 2 Technical Specifications." For these tests the ilowable leak rate per sleeve, in drops per minute, was established by dividing the number of sleeves estimated to be installed in each steam generator into the 0.35 gpm total allowable leak rate per steam generator. "Leak test measurements were based on the number of drops counted during a 10-20 minute period."

The proprietary leak test details and results presented in WCAP 11669 were reviewed. The non-proprietary version of the test 1:; summarized below.

l The following test sequence was used, as describad in the WCAP 11669 report:

"1. Initial leak test: The leak rate was determined at room 0

temperature - 3110 psi and at 600 F - 1600 psi. These tests established the leak rate of the lower joint af ter it has been installed in the steam generator and prior to long-term operation.

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. "2. The specimans were fatigue loaded for 5000 cycles.

"3. The specimens were temperature cycled for 25 cycles.

"4. The specimens were tested at 3110 psi - room temperature and at 1600 psi - 6000F. This established the leak rate after a simulation of 5 years of normal operation.

"S. The specimens were leak tested while being subjected to steam line break conditions.

"6. The specimens were tested at room temperature - 3110 psi and 600 0F

- 1600 psi to determine post accident leak rates."

Westinghouse states: l l

"The leak test results provide verification that the mechanical sleeve exhibits no leakage under what would be considered normal operating  !

conditions and only slight leakage under the umbrella test conditions i

, used. It should be noted that any leakage that was experienced was within the allowable limits. Test conditions were designed to do more than simulate ntual conditions. Rather, they were designed to demonstrata that under extreme accelerated test conditions leakage is  ;

small c. zero, thereby providing assurance that in the actual operating case the sleeves will perform at a zero leakage base. Additionally, by using the same test base for all sleeve designs it was possible to measure consistency in piocess modification and/or small changes in the overall design to facilitate an assessment of their effect on total l sleeve performance."

Westinghause also performed a series of leak rate tests on hybrid expan. ion joint specimens formed out-of-sludge and in-sludge. The tests 0

were at room temperature - 3110 psi and at 600 F - 1600 psi. The leak rates measured during these tests did not exceed the allowable leak rate establisned for a sleeve based on plant Technical Specification leak rates.

A final seriss of leak rate tests were conducted by Westinghouse with fixed / fixed mock ups to simulate the section of the steam generator from the 12 W

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c primary face of the tubesheet to the first support plate. Westinghouse i explains that the term "fixed / fixed" was derived from the fact that the tubes were fixed at these two locations, simulating a dented tube. Inconel 600 sleeves were installed in the tubes of the mock-up and leak tests of the l lower joint and upper hybrid expansion joint were conducted. The test  ;

results at both room temperature and 3500 F and 425 0 F indicated leak rates I within allowable limits.

3.5.2 balytical Verificati_gn The WCAP 11669 report provides analytical evaluation to verify the sleeving repair process tests and results. )

l 3.5.2.1 Structural Analysis The WCAP 11669 report includes a structural evaluation of the sleeve and tube assembly with a hybrid expansion joint and a sleeve with the larger I of the two sleeve lengths proposed for use in the Zion units. The I

structural analyses were performed to demonstrate compliance with the requirements of the ASME Boiler and Pressure Vessel Code,Section III, l Subsection NB, 1983 Edition. The material used in the evaluation is Inconel 690 covered by the ASME Code Case N-20 and the tube material is Inconel 600.

l All taterial properties used in the analyses were as specified in the ASME Boiler and Pressure Vessal Code,Section III, Appendix 1 and related Code ,

l Cases.

The following summary of the structure analysis was provided by Westinghouse.

"The analyses include primary stress intensity evaluations, maximum

! range of stress intensity evaluations, and fatigue evaluations for various mechanical and thermal conditions. The critical portions of the sleeve-tube assembly are two joints, the upper and lower HEJ, and straight sections of the sleeve and tube between the two joints. Past analytical experience indicates that upper joint stresses are more limiting than lower joint stresses. The finite element model developed contains both upper and lower joints. A detailed stress evaluation of the ... sleeve was performed for the upper joint only. The tolerances 13 O. - _,

used in developing the models were such that the maximum sleeve outside diameter was evaluated in combination with the minimum sleeve wall thickness. This allowed the maximum stress levels to be developed in the roll transition regions.

, "Structural analysis of the sleeve-tube assembly included finite element model development, thermal, pressure stress and thermal stress calculations, primary and primary plus secondary stress evaluation, and fatigue evaluation for various mechanical and thermal conditions which r

envelop the loading conditions specified by the appropriate Design and Equipment Specifications. Two computer programs, WECAN and WECEVAL,

, are used in structural analysis of the sleeved tubes.

"The WECAN program performs thermal and stress analysis of the primary structure. Thermal analysis provides the temperatare distribution required for thermal stress calculations. Thermal stress calculations are performed for fixed times under thermal transients. These times for the total pressure and thermal analysis are chosen for the anticipated maximum and minimum total stresses in critical regions of t

the structure. Total stress distribution is determined by combining i the pressure and thermal stress results.

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' Total stress calculations, as well as stress evaluations, are carried out by the WECEVAL computer program, a multi-purpose code, which performs ASME Code,Section III, Subsection NB stress evaluations.

"At any given point or section of the model, the program WECEVAL was used to determine the total stress distribution per the Subsection NB requirements. This stress was categorized into membrane, linear bending, and non-linear components which were compared to the Subsection NB allowables. In addition, complete transient histories at given locations on the model were used to calculate the total cumulative fatigue usage factor per ASME. Code Paragraph NB-3216.2. For the fatigue evaluation, the effect of local discontinuities was considered.

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"The following is a summary of the results of the analysis:

"All primary stress intensities for the sleeved tube assembly are well within all allowable ASME Code limits.

"The requirements of the ASME Code Paragraph NB-3222.2 were met at all locations for the maximum range of stress intensity values for the sleeved assemblies. Based on the sleeve design criteria, the fatigue analysis considered a design life objective of 40 years for the sleeved tube assemblies.

"All of the cumulative usage factors art below the allowable value of 1.0 specified in the ASME Code."

3.5.2.2 Tube Vibration Analyses l

Westinghouse provided the following information about the tube i vibration analysis:

6 "Analytical assessments have been performed to predict nodal natural l frequencies and related dyn&mic bending stresses attributed to flow-induced vibration for sleeved tubes. The purpose of the assessment was to evaluate the effect on the natural frequencies, amplitude of vibration, and bending stress due to installation of various lengths of sleeves. It was found that none of these parameters is adversely affected.

"Since the level of stress is significantly below the endurance limit for the tube material and higher natural frequencies result from the use of a sleeve / tube versus and unsleeved tube, the sleeving j modification does not contribute to cyclic fatigue." l l

3.5.2.3 Effect of Flow-Slot Houralassino Analyses Westinghouse also provided information about several areas of special 6- consideration that were included in its analysis as summarized below.

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"Along the tube lane the tube support plate has long rectangular flow-slots that have the pctential to deform into an hourglass shape with significant denting. The effect of hourglassing is to move the neighboring tubes laterally inward to the tube lane from their initial positions. The maximum bending would occur on the innermost row of tubes in the center of the flow-slots.

i "The effect of hourglassing induced bending stresses on maximum range of stress intensity and fatigue usage factor of the sleeve were calculated. Taking into account the hourglassing induced thermal stress, the largest value of maximum stress intensity would be below the 79.80 ksi allowable, the fatigue usage factor is negligible."

3.5.2.4 Allowable Sleeve Dearadation Analyses Westinghouse states: "Minimum required sleeve wall thicknesses, rt , to sustain normal and accident condition loads were calculated in accordance with the guidelines of Regulatory Guide 1.121. In this evaluation, the J surrounding tube wall was assumed to be completely degraded; that is, no design credit was taken for the residual strength of the tube." The R.G.

l 1.121 requirements state the minimum sleeve wall thickness must meet s factor of safety of three during normal operation and the maximum primary-to-secondary pressure differential that occurs during a steam line break (SLB) accident.

The minimum required sleeve wall thickness was established, which is then used to calculate the plugging limit (see Section 3.5.2.5) with confirmation of leak-before-break. This calculation was described by Westinghouse as sunnarized below.

, "The leak-before-break evaluatic. for the sleeve is based on leak rate and burst pressure test data obtained on 7/8 inch OD x 0.050 inch wall and 11/16 inch OD x 0.040 inch wall cracked tubing with various amounts of thinning simulated by machining on the tube 00. The margins to burst during a postulated SLB accident condition are a function of the

, mean radius to thickness ratio, based on a permissible leak rate of 0.35 gpm due to a normal operating pressure differential of 1530 psi.

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"Using a mean radius to thickness factor of 9.5 for the nominal sleeve, l the current Technical Specification allowable leak rate of 0.35 gpm, a l SLB pressure differatial of 2560 psi, and the nominal leak and nominal  ;

burst curves, a 29.8.; margin exists between the burst crack length and l the leak crack length. For a sleeve thinned 51% through-wall over a 1.0 inch axial length, a 24.8% margin to burst is demonstrated. Thus, l the leak-before-break behavior is confirmed for unthinned and thinned conditions."

3.5.2.5 Pluaaina limit Determination i

Westinghouse then describes how the plugging limit was determined.

'To maintain sufficient tube integrity, allowable levels of tube wall l degradation referred to as plugging limits are established. This determination was performed in accordance with Regulatory Guide 1.121.

Tubes which have eddy current indications of degradation in excess of l the plugging limits must be repaired or plugged. The information 1 l provided in WCAP 11669 defines the portion of the tube and the sleeve for which indications of wall degradations must be evaluated. This information can be summarized as follows:

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"Indications of degradation in the entire length of the sleeve must be evaluated against the sleeve plugging limit.

"Indications of tube degradation of any type including a complete guillotine break in the tube between the bottom of the upper joint and the top of the lower roll expansion does not require that the tube be removed from service.

l r "The tube plugging limit continues to apply to the portion of the i tube in the upper joint and in the lower roll expansion. As noted above the sleeve plugging limit applies to these areas also.

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l "The tube plugging limit continues to apply to that portion of the l tube above the top of the upper joint."

In addition the following part of the proposed amendment to the Zion Technical Specifications that addresses the plugging limit for sleeves was 17

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i provided by the Comonwealth Edison Company in an attachment to its letter of December 24, 1987.

"Westinahouse Electric Corooration Sleeves

"d. for the area of the tube behind the sleeve, above and including the upper joint, any imperfection produced by degradation shall be plugged unless it can be clearly demonstrated by a qualified NDE technique that the imperfection is < 40% of the nominal wall thickness.

"e. for the entire 1er.gth of the sleeve, any imperfection 141% of the I nominal sleeve thickness shall be plugged.

"f. for the c ea of the tube behind the lower joint, any imperfection

! produced by degradation shall be plugged unless it can be clearly demonstrated by a qualified NDE technique that the imperfection is

, < 40% of the nominal wall thickness."

i

, Comonwealth Edison Company's proposed plugging limit for degraded i sleeves (141%) includes a 10% operational allowance for eddy current I

, uncertainty and postulated degradation between inspections, j 3.6 Eddy Current Testino The WCAP 11669 report states that addy current irspections are routinely carried out on the Zion steam generators in accordance with the plant's Technical Specifications. "The purpose of these inspections is to detect at any early stage tube degradation that may have occurred during

plant operation so that corrective action can be tiken to minimize further degradation and reduce the potential for significant primary-to-secondary j leakage."

l Sleeving introduces special inspection problems in the eddy current testing of the sleeve-tube ombination. First a method confirming that the joints meet critical process dimensions is required. Secondly, it must be l shown that the tube / sleeve assembly is capable of being evaluated through subsequent routine in-service in:pection.

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The following summary was provided in the WCAP 11669 report of the eddy current inspection approach for Zion unit sleeve inspections.

'The inspection for degradation of the tube / sleeve assembly has typically been performed using crosswound coil probes operated with a

multifrequency excitation. For the straight length regions of the tube / sleeve assembly, the inspection of the sleeve and tube is I consistent with normal tubing inspections. In tube /siseve assembly I

joint regions, data evaluation becomes more complex.

L "The detection and quantification of the degradation at the transition regions of the sleeve / tube assembly depends on the signal-to-noise ratio between the degradation responta and the transition response. As a general rule, lower frequencies tend to suppress the transition signal relative to the degradation signal at the expense of the ability

! to quantify. Similarly, the inspection of the tube through the sleeve requires the use of low frequencies to achieve detection wiih an associated loss in quantification. Thus, the search for an optimum f eddy current inspection represents a trade-off between detection and quantification.

}

"In the regions of the parent tube above the sleeve, conventional l bobbin coil or crosswound coil inspections will be used. However, since the diameter of the sleeve is smaller than that of the tube, the fill factor of a probe inserted through the sleeve may result in a decreased detection capability for tubing degradation. Thus, it may be necessary to inspect the unsleeved portion of the tube above the sleeve by inserting a standard size probe through the U bend from the unsleeved leg of the tube.

l "For the tube sleeve combination, the use of the crosswound probe, coupled with a multifrequency mixina technique for further reduction of the remaining noise signals significantly reduces the interference from all discontinuities (e.g. transition) which have 360-degree symmetry, providing improved visibility for discrete disuontinuities. In the laboratory this technique can be used to detect outer diameter tube wall penetrations with acceptable signal-to-noise ratios at the 19 W

m. _ - - . - . . - - -

O transitions when the volume of metal removed is equivalent to the ASME calibration standard. -

"For inspection of the region at the top end of the sleeve, the transition response signal-to-noise ratio is about a factor of four less sensitive than that of the expansions. Some improvement has been

, gained by tapering the wall thickness at the top end of the sleeve.

This reduces the end-of-sleeve signal by a factor of approximately two.

The crosswound coil, however, again significantly reduces the response of the sleeve end, and degradation at the top end of the sleeve / tube l assembly can be detected."  !

3.7 Inservice Insoection Plan For Sleeved Tubes The WCAP 11669 report addresses the current NRC requirements to perform periodic inspections of the supplemental pressure boundary. " This new g pressure boundary consists of the sleeve with a joint at the primary face of l

) the tubesheet and a joint at the opposite end of the sleeve."

i "The inservice inspection program will consist of the following. Each sleeved tube will be eddy current inspected on completion of installation to obtain a baseline signature to which all subsequent I inspections will be compared. Periodic inspections to monitor sleeve

, wall conditions will be performed in accordance with the inspection section of the plant Technical Specifications."

4. EVALUATION The evaluation of the amendment proposed by Commonwealth Edison Company and the WCAP 11669 report prepared by Westinghouse is based upon the following considerations. Steam generator tube sleeving is a repair technique that is an alternative to removing defective or degraded tubes from service by plugging. Sleeves are designed to span a defective or degraded region of a steam generator tube and maintain the steam generator

( tubing primary-to-secondary pressure boundary under normal and accident conditions. A successful sleeving system must provide a corrosion resistant sleeve material, and structural integrity and leak tightness of the sleeved tube.

20

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. l The analysis and evaluation supports the use of Inconel 690 TT sleeves as a significant improvement over the Inconel 600 MA material used in the original steam generator tubing. Corrosion tests conducted under Electric Power Research Institute (EPRI) sponsorship confirm WCAP 11669 test results regarding the improved corrosion resistance of Inconel 690 TT over that of

, Inconel 600 MA. Accelerated stress corrosion tests in caustic and chloride aqueous solutions have also indicated that Inconel 690 TT resists general corrosion in the aggressive environments. Isothermal tests in high purity water have shown that, at normal stress levels, Inconel 690 TT has high t

resistance to intergranular stress corrosion cracking in extended high temperature exposure. EPRI (Reference 1) concludes, as a result of these laboratory corrosion tests, that Inconel 690 TT material should be used for PWR steam generator tubing with all volatile treatment secondary water systems. Inconel 690 is a Code approved material (ASME S8-163), covered by ASME Code Case N-20, and accepted by NRC under Regulatory Guide 1.85 i

(Rev.24, July 1986). The 4RC staff has approved Westinghouse's use of

, Inconel 690 TT tubing for sleeving (Reference 9) and in replacement steam

generators modified since 1984 (Reference 7).

[ Corrosion test results presented in WCAP 11669 confirm that the expansion processes used to form the upper and lower sleeve-to-tube joints, while mechanically working the sleeve and tube, did not impose sufficiently j high tensile stresses as to promote stress corrosion cracking in the tube / sleeve samples used in the tests.

P The analyses and qualification test results presented in WCAP 11669 are acceptable and verify the leak resistance of the upper and lower sleeve-to-tube joints and the structural strength of the sleeves under normal and accident conditions. The analyses using Regulatory Guide 1.121 to establish the minimum acceptable sleeve wall thickness necessary for continued service are acceptable. However, the proposed plugging limit for sleeves (2 41%) is not acceptable since the staff does not consider a 10% operational allowance for eddy current uncertainty and operational degradation conservative enough. A 20% operational allowance is acceptable to the staff.

The eddy current techniques proposed for sleeve inspections are acceptable providing more advanced state-of-the-art techniques that are developed and recognized industry wide are used when they become available.

21

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The eddy current inspection program for installed sleeves and the inservice inspection plan for sleeved tubes after periods of operation are also acceptable to the staff.

5. CONCLUSIONS l t

The following conclusions are based upon this review and evaluation of the information and data presented in WCAP 11669 "Zion Units 1 and 2 - Steam Generator Sleeving Report (Mechanical Sleeves)" dated December 1987. It has been concluded that the request by Commonwealth Edison Company for a proposed amendment to Facility License Nos. DPR-39 and DPR-48, Appendix A, Section 43, Reactor Coolant System that will allow Zion's steam generator tubes to be repaired, if needed, by utilizing Westinghouse Electric Corporation mechanical sleeving methodology is acceptable.

l The acceptance of the proposed amendment is contingent upon the licensee's agreement to modify the plugging limit for sleeves from 2 41% to f 1 31% to accommodate a more conservative operational allowance for eddy current uncertainty and any degradation between inspections.

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REFERENCES

1. EPRI- Steam Generator Owners' Group "Steam Generator Reference Book,"

May 1, 1985.

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2. ASME Boiler and Pressure Vessel Code SB-163,Section III through Winter 1983 Addenda.

3. ASME Boiler and Pressure Vessel Code, Code Cases, Code Case N-20, 1983 Edition, July 1, 1983.

l

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4. Regulatory Guide 1.85 (Rev. 24, July 1986).

t

5. December 24, 1987 letter from P.C. LeBlond to NRC with 5 attachments.

(Attachment 1 is the proposed amendment, Attachment 2 is a brief description of the amendment and technical justification, Attachment 3 covers the significant hazards consideration, Attachment 4 contains

. copies of the proprietary and non-proprietary version of WCAP 11669, and Attachment 5 is an affidavit of proprietary information signed by i Westinghouse.)

6. Regulatory Guide 1.121, Plugging Margin.

, 7. NUREG Draft, Steam Generator Operating Experience Update, 1984-1986.

8. Regulatory Guide 1.83, Steam Generator Tube Inspectability.

9. Farley 1&2, September 1987 SER; D.C. Cook,

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