ML20202J799
| ML20202J799 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/30/1998 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC. |
| To: | |
| Shared Package | |
| ML20136E819 | List: |
| References | |
| CEN-629-NP, CEN-629-NP-R03, CEN-629-NP-R3, NUDOCS 9902090249 | |
| Download: ML20202J799 (150) | |
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COMBUSTION ENGINEERING, INC.
CEN-629-NP Revision 03-NP COMBUSTION ENGINEERING, INC.
September 1998 Renair of Westinghnuse Series 44 and 51 Steam Generator Tubes Using Iraktight Sleeves FINAL REPORT Combustion Engineering, Inc.
Nuclear Operations Windsor, Connecticut
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PDR ADOCK 05000282 P
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ABB Combustion Enginecting Nuclear Operatims l
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I i-ABSTRACT t
. A technique is presented for repairing degraded steam generator tubes in pressurized water l
reactor Nuclear Steam Supply. Systems (NSSS). The technique described alleviates the need l
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for plugging steam generator tubes which have become corroded or are otherwise considered i
to have lost structural capability. The technique consists of installing a thermally treated Alloy 690 sleeve which spans the section or sections of the original steam generator tube which j
requires repair. The sleeve is welded to the tube near each end of the sleeve for repairs at the tube support plates or welded at the upper end and lower end or welded at the upper end and hard rolled at the lower end for repairs to the steam generator tube in the tubesheet region.
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i This report details analyses and testing performed to verify the adequacy of repair sleeves for l'
installation in a 7/8 inch O.D. nuclear steam generator tube. These verifications show tube l
sleeving to be an acceptable repair technigtw.
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Report No. CEN-629-NP, Revision 03-NP Page i A
ABB Combustion Engineering Nuclear Operations TABI E OF CONTENTS Section Iitle Eage
1.0 INTRODUCTION
1-1 1.1 PURPOSE 1-1
1.2 BACKGROUND
1-2 1.3 ACRONYMS 1-2 2.0
SUMMARY
AND CONCLUSIONS 2-1 3.0 ACCEPTANCE CRITERIA 3-1 4.0 DESIGN DESCRIPTION OF SLEEVES AND INSTALLATION 4-1 EQUIPMENT j
4.1 SLEEVE DESIGN DESCRIPTION 4-1 4.2 SLEEVE MATERIAL SELECTION 4-1
' 4.3 SLEEVE-TUBE ASSEMBLY 4-2 4.4 PLUGGING OF A DEFECTIVE SLEEVED TUBE 4-4 4.5 SLEEVE INSTALLATION EQUIPMENT 4-4 4.6 ALARA CONSIDERATIONS 4-10
4.7 REFERENCES
FOR SECTION 4.0 4-10 5.0 SLEEVE EXAMINATION PROGRAM 5-1 5.1 ULTRASONIC INSPECTION 5-2 5.2 EDDY CURRENT INSPECTION 5-4 5.3 VISUAL INSPECTION 5-6
5.4 REFERENCES
FOR SECTION 5.0 5-7 Report No. CEN-629-NP, Revision 03-NP Pageii
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1 ABB Combustion Engineering Nuclear Operations T IRT OF TABIRR i
NIL IiLic Eage lI-1 ACRONYMS USED IN REPORT 1-3 2-1 INSTALLATIONS OF ABB CENO WELDED SLEEVE 2-3 j
3-1 REPAIR SLEEVING CRITERIA 3-2 l
L 5-1 ACRONYMS USED IN ET ANALYSIS 5-8 6-1 STEAM GENERATOR TUBE SLEEVE CORROSION TESTS 6-2 6-2
. SLEEVE-TUBE CAPSULE SCC TESTS 6-4 l
i 6-3
- SECONDARY SIDE STEAM GENERATOR TUBE SLEEVE 6-8 CAPSULE TESTS i
l 7-1 SLEEVE-TUBE ASSEMBLY MECHANICAL TESTING-7-5 RESULTS l
l 8-1
SUMMARY
OF SLEEVE AND WELD ANALYSIS SIGNIFICANT 8-4 RESULTS i
8-2
SUMMARY
OF LOWER JOINT (WELDED AND ROLLED) 8-6 L
DESIGN, ANALYSIS, AND TEST RESULTS
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l 8-3A' 30 INCH SLEEVE - AXIAL MEMBER PHYSICAL PROPERTIES -
8-15 l
OR OPERATING STEAM GENERATOR j
8-3B 30 INCH SLEEVE - AXIAL MEMBER PHYSICAL PROPERTIES FOR 8-16 i
" WORST" CASE ENVELOPMENT l
84A' 30 INCH SLEEVE - AXIAL LOADS IN SLEEVE WITH TUBE NOI 8-17 LOCKED INTO TUBE SUPPORT FOR OPERATING STEAM l
GENERATOR l
8-4B 30 INCH SLEEVE - AXIAL LOADS IN SLEEVE WITH TUBE NOI 8-18 LOCKED INTO TUBE SUPPORT FOR " WORST" CASE ENVELOPMENT i
8-5A 30 INCH SLEEVE - AXIAL LOADS IN SLEEVE WITH TUBE 8-19 LOCKED INTO TUBE SUPPORT FOR OPERATING STEAM GENERATOR:
i Report No. CEN429-NP, Revision 03-NP Page v f
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m ABB Combustion Engineering Nuclear Operations I IRT OF TABI RR (Continued)
Na Iitle Eage 8-5B 30 INCH SLEEVE - AXIAL LOADS IN SLEEVE WITH TUBE 8-20 i
LorKFn INTO TUBE SUPPORT FOR " WORST" CASE ENVELOPMENT i
8-6 UPPER SLEEVE WELD-TRANSIENTS CONSIDERED 8-26
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8-7 LOWER SLEEVE SECTION-TRANSIENTS CONSIDERED 8-27 8A-1A STRESS RESULTS 100% STEADY STATE 8A-5 8A-1B '
STRESS RESULTS 0% STEADY STATE 8A-6 8A-IC STRESS RESULTS REACTOR TRIP 8A-7 i
8A-2 FATIGUE EVALUATION AT WORST LOCATION 8A-8 I
8A-3A STRESS RESULTS 100% STEADY STATE (0.020" Weld) 8A-10 8A-3B STRESS RESULTS 0% STEADY STATE (0.020" Weld) 8A-11 8A-3C STRESS RESULTS REACTOR TRIP (0.020" Weld) 8A-12 l
8A-4A RANGE OF STRESS AT WORST LOCATIONS (0.020" Weld) 8A-13 8A-4B FATIGUE EVALUATION AT WORST LOCATIONS (0.020" Weld) 8A-14 8B-1 STRESS RESULTS 8B-7
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- FATIGUE EVALUATION 8B-8 9-1 0.875 O.D. SLEEVED TUBE PWHT DATA 9-8 i
9-2 0.750" O.D. SLEEVED TUBE PWHT DATA 9-8 TUBES LOCKED AT ALL SUPPORTS
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9-3 ABB CENO S/G SLEEVE OPERATING HISTORY 9-9 10-1 TYPICAL SLEEVE TO PLUG EQUIVALENCY RATIO 10-2
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Report No. CEN-629-NP, Revision 03-NP'
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t I IST OF FIGURFR (Continued) l Na Iitle P_ age i
5-1 NDE PROCESS FLOW CHART 5-9 l
l 5-2 ET PROCESS FLOW CHART 5-10 5-3 UT B SCAN - ACCEPTABLE 5-11
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l 5-4 UT B SCAN - REJECTABLE 5-12
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5-5 UT CALIBRATION STANDARD 5-13 6-1 PURE WATER CORROSION TEST SPECIMEN 6-11 6-2 ATS WELD CAPSULE TEST SPECIMEN 6-12 1
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6-3 TSP WELD CAPSULE TEST SPECIMEN 6-13 I
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6-4 CAUSTIC CORROSION AUTOCLAVE TEST SPECIMEN 6-14 i
I 8-1 WELDED SLEEVE / TUBE ASSEMBLY 8-32 8-2 SYSTEM SCHEMATIC FOR OPERATING STEAM GENERATOR 8-33 t
3-3 SYSTEM SCHEMATIC FOR " WORST" CASE ENVELOPMENT 8-34 8-4 STIFFNESS MODEL OF SLEEVE AND LOWER TUBE 8-35 I
8-5 STIFFNESS MODEL OF UPPER TUBE AND SURROUNDING TUBES 8-36 I
i 8-6 FINITE ELEMENT MODEL OF UPPER TUBE WELD 8-37 i
8-7 FINITE ELEMENT MODEL OF LOWER STUB WELD 8-38 l
t 8-8 TUBESHEET PERFORATED PLATE LIGAMENT STRESSES 8-39 3
8A-1 NODE AND STRESS CUT IDENTIFICATION 8A-4 8A-2 NODE AND STRESS CUT IDENTIFICATION FOR 20 MIL WELD 8A-9 l
8B-1 LOWER STUB WELD MODEL (HOT STANDBY) 8B-3 l
Report No. CEN-629-NP, Revision 03-NP Page viii l
l ABB Combustion Engineering Nuclear Operations I IST OF FIGURES (Continued)
I HL Iitic Eage 8B-2 LOWER STUB WELD MODEL (FULL POWER & THERMAL LOAD) 8B-4 8B-3 LOWER STUB WELD MODEL (REACTOR TRIP & THERMAL LOAD)8B-5 8B-4 LOWER STUB WELD MODEL (SECONDARY LEAK TEST) 8B-6 9-1 0.875" O.D. LOCKED TUBE TEST, TEST MOCKUP 9-10 t
9-2 0.875" O.D. LOCKED TUBE TEST, 9-11 TEMPERATURE AND AXIAL LOAD PROFILE l
9-3 0.750" O.D. LOCKED TUBE TEST, TEST MOCKUP 9-12 9-4 0.750" O. D. TYPICAL TEMPERATURE PROFILES 9-13 l
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f Report No. CEN-629-NP, Revision 03-NP Page ix
ABB Combustion Engineering Nuclear Operations FOREWORD As noted in the topical report, CEN429-P, " Repair of Westinghouse Series 44 and 51 Steam Generator Tubes Using Leaktight Sleeves", the tooling and methods described represent the current technology implemented for sleeve installation and inspection. As technological advances are made in sleeve installation and/or inspection techniques, the new tooling and/or processes may be utilized after they have been laboratory verified to provide improved sleeve installation methods, or after a suitable qualification program has ' demonstrated improved performance. Such advances / improvements may be implemented provided that they do not involve alternative joining technology or alternative sleeve material, and provided that the
.j 10CFR50.59 process has demonstrated that no unreviewed safety question will be created. The 10CFR50.59 process will be performed under the licencee's program.
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i Report No. CEN429-NP, Revision 03-NP Page x
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1.0 INTRODUCTION
1.1 PURPOSE j
i The purpose of this report is to provide information sufficient to support a technical l
specification change allowing installation of repair sleeves in Westinghouse designed l
Series 44 and 51 steam generators. This report demonstrates that reactor operation -
with sleeves installed in the steam generator tubes will not increase the probability or consequence of a postulated accident condition previously evaluated. Also it will not create the possibility.of a new or different kind of accident and will'not reduce the existing margin of safety.
I ABB Combustion Engineering (ABB-CE) provides three types of leaktight sleeves for repair of 7/8 inch O.D. steam generator tubes with partial depth rolled tubesheet 4
joints or full depth rolled or expanded tubesheet joints. The first two types span the parent steam generator tube within the tubesheet. One type of tubesheet sleeve is welded to the tube near both the upper and lower end of the sleeve. The second type i
of tubesheet sleeve is welded near the upper end and hard rolled into the tube within the steam generator tubesheet. 'A variation on these designs involves the use of a pre-curved sleeve to install a tubesheet sleeve at the periphery of the tube bundle. The steam generator tube with the installed sleeve meets the structural requirements of tubes which are not degraded.
The third type of sleeve spans degraded areas of the steam generator tube at a tube support or in a free span section of tube. This leaktight sleeve is welded to the steam -
generator tube near each end of the sleeve. The steam generator tube with the installed welded sleeve meets the structural requirements of tubes which are not i
I degraded.
Design criteria for all types of sleeves were prepared to ensure that all design and j
licensing requirements are considered. Extensive analyses and testing have been j
. performed on the sleeve and sleeve to tube joints to demonstrate that the design criteria are met.
The effect of sleeve installation on steam generator heat removal capability. and system flow rate are. discussed in this report. Heat removal capability and system j
flow rate was considered for installation of one to three sleeves in a steam generator j
-tube.
i Plugs will 'be installed 'if ' sleeve installation is not successful or if there is l
unwymble degradation of a sleeve or sleeved steam generator tube. Mechanical l
I sleeve plugs or standard mechanical steam generator tube plugs may be used to take a j
sleeved tube out of service.
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1 Report No. CEN-629-NP, Revision 03-NP Page 1-1 f
r ABB Combustion Engineering Nuclear Operations
1.2 BACKGROUND
The operation of Pressurized Water Reactor (PWR) steam generators has in some instances, resulted in localized corrosive attack on the inside (primary side) or outside
-(secondary side) of the steam generator tubing. This corrosive attack results in a reduction in steam generator tube wall thickness. Steam generator tubing has been j
designed with considerable margin between the actual wall thickness and the wall thickness required to meet structural requirements. Thus it has not been necessary to i
take corrective action unless structural limits were being approached.
Historically, the corrective action taken when steam generator tube wall degradation has been severe has been to install plugs at the inlet and outlet of the steam generator tube when the reduction in wall thickness reached a calculated value referred to as a l
plugging criteria. Eddy current (ET) examination has been used to measure steam generator tubing degradation and the tube plugging criteria accounts for ET measurement uncertainty.
Installation of steam generator tube plugs removes the heat transfer surface of the plugged tube from service and leads to a reduction in the primary coolant flow rate r
available for core cooling. Installation of welded and/or welded and hard rolled steam generator sleeves does not significantly affect the heat transfer removal capability' of the tube being sleeved and a large number of sleeves can be installed l
without significantly affecting primary flow rate 1.3 -
ACRONYMS Table 1-1 (along with Table 5-1) contains a list of the acronyms used throughout this report.
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Report No. CEN-629-NP, Revision 03-NP Page 1-2 l
r ABB Combustion Engineering Nuclear Operations.
TABLE 1-1 L
ACRONYMS USED IN RFPORT
+ POINT: + Point
- 3 ATS: Above Tubesheet EFPH: Effective Full Power Hours EFPY: Effective Full Power l' ears ET: Eddy Current Testing ETZ: Expansion / Roll Transition Zone EW: Edge Weld FDTS: Full Depth Tubesheet LOF: Lack of Fusion PWHT: Postweld Heat Treatment 4
TS: Tube Support UT: Ultrasonic Testing VT: Visual Testing WTS: Welded Tubesheet
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f Report No. CEN-629-NP, Revision 03-NP Page 1-3
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- 2.0
SUMMARY
AND CONCLUSIONS i
l The sleeve dimensions, materials and joints were designed to the applicable ASME j
Boiler and Pressure Vessel Code.
An extensive analysis and test program was j
undertaken to prove the adequacy of both the welded / welded and welded /hard rolled l
sleeve. This program determined the effect of normal operating and postulated i
accident conditions on the sleeve-tube assembly, as well as the adequacy of the assembly to perform its intended function. The proposed sleeving provides for a substitution in kind for a portion of a steam generator tube. The proposed change has j
no significant effect on the configuration of the plant, and the change does not affect
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the way in which the plant is operated. Design criteria were established prior to i
performing the analysis and test program which, if met, would prove that these sleeve i
types are an acceptable repair technique. This set of criteria conformed to the stress
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limits and margins of safety of Section III of the ASME B&PV Code. The safety l
factors of 3 for normal operating conditions and 1.5 for accident conditions were l
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applied. Based upon the results of the analytical and test programs described in this report, these sleeve types fulfill their intended function as leaktight structural i
members and meet or exceed all the established design criteria.
Evaluation of the sleeved tubes indicates no detrimental effects on the sleeve-tube assembly resulting from reactor system flow, coolant chemistries, or thermal and l
pressure conditions. Structural analyses of the sleeve-tube assembly, using the demonstrated margins of safety, have established its integrity under normal and accident conditions. The stmetural analyses have been performed for sleeves which span the tube to a maximum length of 30.0 inches, sleeves which span a tube support or free span length of tube with a length of 9.0 inches and a combination of the sleeve types. The stmetural analyses performed are applicable to shorter tubesheet
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i and tube support sleeves. The analyses for the different sleeve types and lengths are given in Section 8.
Mechanical testing using ASME code stress allowables has been performed to support the analyses. Corrosion tests of typical sleeve-tube assemblies have been completed and reveal no evidence of sleeve or tube corrosion considered detrimental under J
anticipated service conditions.
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Based upon the testing and analyses performed, the proposed sleeves do not result in a significant increase in the probability of occurren::e or consequence of an accident
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previously evaluated, create the possibility for a new or different kind of accident, or result in a significant reduction in a margin of safety.
J Welding development has been performed on clean tubing, dirty tubing which has j
1 been taken from pot boiler tests and contaminated tubing taken from k number of steam generators. ABB-CE installed their first welded sleeves in a demonstration I
program at Ringhals Unit 2 in May 1984. ABB-CE's sleeving history is shown in Table 2-1. The success rate for all installed sleeves is 98%. Since 1985, no sleeve Report No. CEN-629-NP, Revision 03-NP Page 2-1 1
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ABB Combustion Engineering Nuclear Operations which has been accepted based on NDE has been removed from service due to service induced degradation.
If a steam generator tube which has been sleeved is found to require plugging to remove it from service a mechanical sleeve plug or a standard mechanical steam i
generator tube plug can be installed. No discussion or evaluation of the mechanical tube or sleeve plug is provided as part of this document.
In conclusion, steam generator tube repair by installation of any of the three types of sleeves described herein is establishui as an acceptable method.
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i Report No. CEN-629-NP, Revision 03-NP Page 2-2
t ABB Combustion Engineering Nuclear Operations t
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TABLE 2-1 0'
2-INSTALLATIONS OF ABB CENO WELDED SLEEVE PLANT
.DATE QUANTITY TYPE
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i Prairie Island 1 11/97 222 WTS
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Kewaunee 5/97 428 WTS t
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Zion 2 9/%
237 WTS i
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273-TS i
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'188 ETZ
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Byron l' 4/%
3527 ETZ
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Prairie Island 1 2/%
253 WTS l
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-ANO2 10/95 711 ETZ l
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Zion 1 10/95 911 WTS i
Zion 2 1/95 162 WTS i
i KRSKO1 6/93 160 ETZ f
1 14 TS i
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Ginna 4/93 51 WTS l
Zion 2 12/92 172 WTS j
c Prairie Island 1 11/92
_158 WTS i
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Ginna '
4/92 175 WTS 63 Curved WTS Zion 1 4/92 124 WTS Kewaunee 3/92 16 Curved WTS Report No. CEN-629-NP, Revision 03-NP Page 2-3
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ABB Combustion Engineering Nuclear Operations TABLE 2-1 (Continued)
INSTAILATIONS OF ABB CENO WF1DED SI FEVE PLANT DATE QUANTITY TYPE
- Ringhals 3 7/91 46 ETZ 22 TS Ginna 4/90 192 WTS l
48 Curved WTS i
Zion 2 4/90 82 WTS Prairie Island 1 1/90 63 WTS Zion 1 9/89 445 WTS Ginna 4/89 395 WTS 107 Curved WTS Prairie Island 1 9/88 74 WTS Ringhals 2 5/87 571 WTS Prairie Island 1 4/87 27 WTS 1
Ginna 2/87 105 WTS 1
Zion 1 10/86 128 WTS Ringhals 2 5/86 599 WTS Ginna 2/86 36 WTS Ringhals 2 5/85 59 WTS Ringhals 2 5/84 18 WTS
- Straight sleeves unless otherwise noted Report No. CEN-629-NP, Revision 03-NP Page 2-4
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ABB Combustion Engineering Nuclear Operations l
3.0 ACCEPTANCE CRITERIA The objectives of installing sleeves in steam generator tubes are twofold. The sleeve J
must maintain structural integrity of the steam generator tube during normal operating and postulatal accident conditions. Additionally, the sleeve must prevent
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leakage in the event of a through-wall defect in the steam generator tube. Numerous tests _and analyses were performed to demonstrate the capability of the sleeves to perform these functions under normal operating and postulated accident conditions.
l Design and operating conditions used to bound the applicable steam generators are defined as:
i Primary Side:
594*F (operating) 2250 psia (operating)
(Hot Side) 650*F (design) 2560 psia (design)
Secondary Side:
467*F (100% load) 653 psia (100% load) -
l 550*F (design) 1085 psia (design)
Though some plants may operate at higher or lower temperatures and/or pressures, the conditions analyzed represent the greatest temperature and pressure differentials for all operating plants combined. Thus, the loadings calculated in Section 8.0 reflect the maximum possible theoretical values that can be obtained.
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Table 3-1 provides a summary of the criteria established for sleeving in order to j
demonstrate the acceptability of the sleeving techniques. Justification for each of the i
criterion is provided. Results indicating the minimum level with.which the sleeves surpassed the criteria are tabulated. The section of this report describing tests or analyses which verify the characteristics for a particular criterion is referenced in the
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table.
t Repon No. CEN-629-NP, Revision 03-NP Page 3-1 i
ABB Combustion Engineering Nuclear Operations TABLE 3-1 REPAIR SLEEVING CRITERIA
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CRITERION JUSTIFICATION RESULTS SECT 1.
Sleeve is leaktight I2akage between primary Welded sleeve is leaktight 4.0 and secondary side is and is checked using prevented when steam ultrasonic testing, eddy generator tube is breached.
current or visual examination.
2.
Sleeve-tube assembly Sleeve tube assembly Stmetural margins maintained 8.0 functional integrity must meets applicable ASME for all conditions, be maintained for normal Code requirements.
operating and accident.
conditions.
3.
Axial load cycle 200 Bounds thermal cycle No cracking in upper or 7.3 pounds to 1700 pounds for loading from normal lower weld joint. No damage 1000 cycles,200 pounds operating and transient to sleeve or tube.
to 2550 pounds for 1000 cycling.
cycles without weld failure 4.
Pressurization of annulus Prevention of sleeve Assembly collapse at 5200 7.3 between sleeve and tube' failure for through wall psig.
does not collapse sleeve at defect in tube wall.
1500 psig.
5.
Pressurize sleeve to 4800 Factor of safety of three No assembly burst at up to 8.3 psig without bursting.
(3) for normal operating 7800 psig.
(1600 psig delta pressure) conditions.
6.
Exposure of sleeve-tube Sleeve-tube assembly No detectable indication of 6.0 sleeve assembly to various required to function under sleeve orjoint corrosion or prunary and secondary coolant chemistries aggravated tube corrosion.
chemistries without loss of functional integrity.
7.
Non-destructive Periodic examination of ECT technique developed that 5.0 examination of tube and-tubes and sleeves required exceeds EPRI guidelines and sleeve to levels of to verify stmetural Appendix H requirements.
detectability required to adequacy show structural adequacy 8.
Welded sleeve installation Sleeve repair should not System flow rate and heat 10.0 does not significantly reduce power removal transfer capability are not affect system flow rate or capability of reactor or significantly affected.
heat transfer capability of steam generator below the steam generator.
rated value.
Report No. CEN-629-NP, Revision 03-NP Page 3-2
ABB Combustion Engineering Nuclear Operations 4.0 DESIGN DESCRIPTION OF SLEEVES AND INSTALLATION EQUIPMENT f
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4.1 SLEEVE DESIGN DESCRIPTION
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There are three (3) types of sleeves which may be installed in various combinations within a steam generator tube. These sleeves are shown in Figures 4-1 through 4-4.
1 Each sleeve type has a nominal outside diameter of 0.743 inches and a nominal wall i
i thickness of 0.039 inches. The sleeve material is thermally treated Alloy 690. Each sleeve type includes a chamfer at both ends to provide a lead in for equipment used to
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install the sleeve and to facilitate the inspection of the steam generator tube and j
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l The first type of sleeve, shown in Figure 4-1, spans the expansion transition zone and the full length of the tube within the tubesheet. This Full Depth Tubesheet (FDTS) l l
sleeve with a welded lower joint is up to 30.0 inches long and chamfered at the upper end. The lower end of the sleeve is tapered prior to welding. The taper provides a j
tight fit between the sleeve and tube to facilitate welding. The taper also provides t
lead in at the lower end of the sleeve.
The second type of sleeve, shown in Figure 4-2, also spans the full length of the tube l
within the tubesheet. This Full Depth Tubesheet (FDTS) sleeve with a rolled lower l
joint is up to 30.0 inches long and includes a band of nickel and a band of metal oxide on one end. The nickel band improves sealing of the sleeve when the lower i
end is hard rolled into the expanded portion of the parent tube while the rough j
surface of the metal oxide enhances the strength of the mechanicaljoint.
l A variation of these sleeves, the peripheral FDTS sleeve, is shown in Figure 4-3.
l This sleeve is preformed to a radius of curvature of approximately 38 inches over the majority ofits length as part of the manufacturing process. This curvature allows the sleeve to be installed in peripheral tube locations where the primary head to tubesheet l
clearance would not permit the installation of a straight FDTS sleeve. Short straight lengths of approximately four inches on the upper end and two inches on the lower
)
end of the sleeve are provided for installation tooling requirements.
The third type of sleeve, shown in Figure 4-4, spans a tube support. This Tube Support (TS) sleeve is 9 inches in length. The sleeve spans a tube support elevation or can be used on a free span section of the tube. One or two TS sleeves may be used in a tube and may be used in a tube containing a FDTS sleeve.
1 4.2 SLEEVE MATERIAL SELECTION The thermally treated Alloy 690 tubing, from which the sleeves are fabricated, is procured to the requirements of the ASME Boiler and Pressure Vessel Code,Section II SB-163, Code Case N-20. Additional requirements are applied including a limit on Carbon content of 0.015 - 0.025% and a minimum annealing temperature of 1940 F Report No. CEN-629-NP, Revision 03-NP Page 4-1
ABB Combustion Engineering Nuclear Operations i
i (1060"C). The thermal treatment is specified at 1300*F (704*C) to impart greater
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corrosion resistance in potential faulted secondary side environments. The enhanced j
corrosion resistance is achieved in the thermal treatment by insuring the presence of l
grain boundary carbides and by reducing the residual stress level in the tubing.
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During sleeve fabrication for a peripheral sleeve, an intermediate stress relief anneal
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is employed to reduce residual stresses induced during forming of the sleeve to the i
curved shape. This stress relief anneal is carried out at 1300 F for two hours.
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Upon stress relieving, the sleeve is ready for insertion and welding.
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The mechanical propenies of the sleeve material, particularly in the weld areas, are i
j not affected by the forming or stress relieving since the top and bottom portions of
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the sleeve are not curved or restraightened. A four inch length at the top and two i
inch length at the bottom of the sleeve remains undisturbed with respect to bending i
and constitutes "as received ' tubing stock. The minor exception to this statement is the additional two hours at 1300*F the ends are exposed to in sleeve stress relieving
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of the formed or curved configuration. This constitutes a time extension of the thermal treatment time employed at the tube mill which can be viewed as a positive
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exposure. Sleeve expansion, flaring, welding and rolling takes place in sleeve ends that have not been curved and with tooling and processes duplicating those used for straight sleeve installation.
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The principal selection criterbn for the sleeve material was its resistance to stress I
corrosion cracking (SCC) in primary and caustic faulted secondary PWR environments. ABB-CE's justification for selection of this material and condition is j
based on the data contained in Reference 4.7.1.
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4.3 SLEEVE-TUBE ASSEMBLY 1
The installed sleeves are shown in Figures 4-5 through 4-7. The full depth tubesheet (FDTS) sleeves span the tube in the expanded region, the creviced tubesheet region (where applicable), and the lower portion of the tube just above the secondary face of the tubesheet. If defects exist at a tube support as well as within the tubesheet, a FDTS sleeve and a Tube Support (TS) sleeve may be used.
The welded FDTS sleeve, shown in Figure 4-5, is 30 inches in length or shorter. The bottom of the 30 inch sleeve is flush with the bottom of the steam generator tube and extends approximately 8 inches above the secondary face. The welded FDTS sleeve is expanded in the steam generator tube at the upper end in preparation for welding.
A hydraulic expansion tool, as described in Section 4.5.5, is used to expand the sleeve. The sleeve expansion at the upper end results in a slightly expanded steam generator tube which springs back after the expansion is complete resulting in a minimal gap between the sleeve and tube. The expansion holds the sleeve in place for welding.
Report No. CEN-629-NP, Revision 03-NP Page 4-2
I ABB Combustion Engineering Nuclear Operations The lower end of the sleeve is tapered prior to welding. The taper serves to provide l
tight contact with the tube for welding.
A weld is made at the upper end of the sleeve to join it to the steam generator tube.
Upper weld penetration into the steam generator tube is limited to less than 100% of l
the. wall thickness with a minimum average weld height of 0.020 inches at the sleeve-tube interface. The upper sleeve to steam generator tube weld is centered approximately 1-1/4 inches below the end of the sleeve.
If this upper weld is l
defective, it can be repaired by gas tungsten arc remelting of the original weld using j
the same welding procedure parameters as were employed for the original weld. The lower weld is'a full penetration edge weld at the end of the sleeve and tube. If the lower weld is defective, it can also be repair welded.
The welded and rolled FDTS sleeve is also 30 mehes in length or shorter. The l
bottom of this sleeve is flush with the tube end for partial depth rolled tubesheet l
joints. The bottom of this sleeve is located approximately at the tubesheet neutral axis for full depth rolled or expanded tubesheet joints. The upper end of this 30 inch 1
FDTS sleeve is located 8 inches above the tubesheet upper face. This FDTS sleeve is expanded and welded into the steam generator tube at the upper end in the same manner as described above while the lower end of the sleeve is hydraulically i
expanded in preparation for rolling. A hydraulic expansion tool, as described in Section 4.5.5, is used to expand the sleeve.
)
i The weld process, repair weld process and welding operators have been qualified for making upper and lower welds. The weld process and weld operator qualifications are described in Section 9.0.
l 1
The lower end of the welded and rolled FDTS sleeve is rolled into the tube within the tubesheet. The roll is controlled to provide a leaktight structural joint. A roll which does not meet the roll acceptance criteria can be repaired by rerolling at the same location.
The TS sleeve shown in Figure 4-7 is 9 inches in length. It is approximately centered at a tube support plate or eggerate support.
The upper and lower ends are l
hydraulically expanded into the steam generator tube to hold the sleeve in place for welding and to provide the sleeve to tube fit-up necessary for welding.
L Weld penetration into the steam generator tube is limited to less than 100% of the wall thickness and the minimum average weld height is 0.020 inches at the sleeve to tube interface. The upper and lower welds are centemd approximately 1 1/4 inches from the ends of the sleeve. If the weld is defective it can be repaired by gas tungsten arc remelting of the original weld using the same welding procedure parameters employed for the original weld.
Report No. CEN-629-NP, Revision 03-NP Page 4-3
i ABB Combustion Engineering Nuclear Operations When it is considered to be of benefit, a postweld heat treatment of the upper sleeve to tube weld in a tubesheet sleeve and both welds in a tube support sleeve will be added to the sleeve installation process. After the sleeve has been welded into the tube, the weld joints are heated in the range of [
].As described in Reference 4.7.5, this time and temperature combination is sufficient to reduce the level of residual stress in Alloy 600 while minimizing detrimental effects such as grain growth or sensitization. This treatment is similar to that utilized in some operating units to heat treat the tight radius U-bends.
Based on plant specific operating conditions (temperatures, pressures, chemistry, etc) and steam generator tube degradation history, postweld heat treatment may not be of any significant benefit. Five non-postweld heat treated sleeves installed at Ringhals 2 (Westinghouse 51 series steam generator) in 1985 and 1986 were removed in January 1990 and extensively examined. These sleeves (and the parent tubing at the sleeve-l tube joint) which had accumulated up to 920 Effective Full Power Days (EFPD) of I
i service, showed no field service degradation. One non-post weld heat treated sleeve installed in Prairie Island (Westinghouse 51 Series S/G) in 1987 was removed in l
1997. This sleeve, which had accumulated 3358 EFPD of service, showed no field service degradation in the sleeve or parent tube in the ATS weld joint region.
Qualification of the sleeve welding process is in accordance with the procedure described in Section 9.0.
l 4.4 PLUGGING OF A DEFECTIVE SLEEVED TUBE l
l i
I If a sleeve or sleeved tube is found to have an unrepairable defect, the tube can be taken out' of service with mechanical sleeve plugs or standard mechanical steam generator tube plugs installed at both ends of the tube using approved methods. The Regulatory Guide 1.121 analysis for the sleeve is included in Section 8.3.
4.5-SLEEVE INSTALLATION PROCESS AND EQUIPMENT The equipment used for remote installation of sleeves in a steam generator is made up l
of the following basic systems. These systems are:
1.
Remote Controlled Manipulator 2.
Tool Delivery Equipment 3.
Tube Brushing-Cleaning Equipment l
4.
Tube Size Rolling Equipment I
5.
Sleeve Expansion Equipment j
i 6.
Sleeve Welding Equipment j
Report No. CEN-629-NP, Revision 03-NP Page 4-4 l
-,.. -.~
T.
i ABB Combustion Engineering Nuclear Operations 1
7.-
Nondestructive Examination Equipment l
' 8.
Sleeve Rolling / Tapering Equipment
.9.
Sleeve Heat Treatment Equipment
.j These systems, when used together, allow installation of the sleeves without t
personnel entering the steam generator. In this way, personnel exposure to radiation
)
'is held to a minimum.
i The tooling and methods described in the following sections represent the present technology for leaktight sleeve installation. As technological advances are made in sleeve installation, the new tooling and/or processes inay be utilized after they have i
been laboratory verified to provide improved sleeve installation methods.
4.5.1.
Remote Controlled Maninnintnr The remote controlled manipulator (Figure 4-8) serves as a transport vehicle for inspection or repair equipment inside a steam generator primary head.
The manipulator consists of two major components; the manipulator leg and manipulator arm. The manipulator leg is installed between the tubesheet and bottom of the primary head and provides axial (vertical) movement of the arm. The manipulator l
arm is divided into the head arm, probe arm and a swivel arm. Each arm is moved j
independently with encoder position controlled electric motors. The swivel arm j
allows motion for tool alignment in both square pitch and triangular pitch tube arrays.
l Computer control of the manipulator allows the operator to move sleeving tools from l
outside the manway and accurately position them against the tubesheet.
j i
4.5.2 Tooting Deliverv Faniament The purpose of the tooling delivery equipment is to support and vertically position the various tools required for the sleeving operation and to provide controlled rotation to some of the tools. Two different delivery systems may be used for the l
tool delivery. A probe pusher located on the platform outside the steam generator is i
used for tools which do not require controlled rotation. A tool driver located at the tubesheet provides delivery of tooling requiring controlled rotation. The tool driver may also be used to deliver sleeving tools in place of the probe pusher located on the platform.-
.l The probe pusher is a modified Zetec probe pusher or equivalent unit located outside the manway of the steam generator. A flexible conduit extending from the probe l
pusher to an adaptor on the manipulator arm provides the guide path for the tools.
l The guide path adaptor is attached to the end of the manipulator arm by a dovetail
. fitting and manual lock. The drive wheels of the probe pusher deliver the tools to the i
Report No. CEN-629-NP, Revision 03-NP Page 4-5 w
I i
ABB Combustion Engineering Nuclear Operations required elevations within the tube. Where positioning is critical, a hardstop attached to the tool shaft locates the tool relative to the steam generator tube end.
i The tool delivery system for controlled tool rotation consists of two major j
components; a tool mounting plate and a tool driver, shown in Figure 4-9. The tool l
mounting plate is attached to the end of the manipulator arm by a dovetail fitting and manual lock. One or two sets of pneumatically operated fingers are used to draw-up and lock the tool mounting plate to the tubesheet. Proper alignment of the tool l
mounting plate to the tubesheet is assured through the actuation of three sensors i
against the tubesheet. A spring loaded, air pressure release, quick change mount is provided on the face of the tool mounting platform for quick mounting of the tool driver.
l The tool driver attaches to the tool mounting platform with the quick change mount.
I The tool driver includes two double sets of drive wheels and two idler wheels. The drive wheels are powered by electric motors to insert and remove the various j
sleeving tools into and out of the steam generator tube. Vertical positioning of the tools is accomplished by hardstops and/or verified by visual means. Controlled rotation of the weld and non-destructive examination (NDE) tools is provided by an electric motor which rotates the tool driver relative to the tool mounting platform.
t Modifications to the mounting configuration allows the use of the tool driver directly off of the manipulator arm, bypassing tle need for the mounting plate.
Either approach results in identical tube accessibility throughout the steam generator.
i 4.5.3 Tube Bmshine-ciennina Fnninment Prior to sleeve installation, the tube I.D. is mechanically cleaned with a centrifugal j
wire brush. This operation is performed using a tool similar to that shown in Figure i
4-10. A motor rotates the tool head as it is inserted in the end of the tube. At the appropriate elevation, the tool is reciprocated at approxirnately one inch per second over each FDTS and TS sleeve joint area. In order to remove any oxide layer from the tube I.D. in preparation for welding, the brush head must rotate at a minimum speed of 2000 rpm for at least two minutes over each sleeve joint area. As new l
I technology becomes available, alternate cleaning methods may be employed.
Two methods are available to remove brushing fines and loose material prior to j
sleeve installation. A lint free cotton swab is used to dry wipe the brushed areas of the steam generator tube. The swab is rotated at a speed up to 750 rpm and is reciprocated twice over each cleaned section of the tube.
Alternatively, a high l
pressure blowing system may be used to remove any loose material prior to sleeve l
installation.
I i
Report No. CEN-629-NP, Revision 03-NP Page 4-6 l
l
m
' ABB Combustion Engineering Nuclear Operations 4.5.4 Tube Rollino Foninment The tube brushing-cleaning tool described in Section 4.5.3 contains a go/no-go gage which is used to determine whether or not sleeve insertion restrictions exist at the tube end I.D. Size rolling (to a nominal tube I.D.) may be used, if necessary, to remove any tube end restrictions to facilitate sleeve installation. The tube roller is powered by the air operated rolling tool described in Section 4.5.9.
l 4.5.5 Sleeve Fvnansinn Fnninment The sleeve expansion equipment is used to provide the required sleeve-tube fit-up prior to welding the upper FDTS weld or either TS weld.
The sleeve is located on the sleeve expansion tool for positioning within the steam generator tube. The expansion tool functions to guide the sleeve into the tube and install the sleeve to the selected elevation within the tube.
A tool hardstop is provided for proper sleeve. vertical positioning. Once the sleeve is at the proper elevation within the steam generator tube, it is hydraulically expanded.
The expansion tool, shown in Figure 4-11, consists of a mandrel and two bladders which contain the demineralized water used as the pressurization fluid. The sleeve is located over the two bladders prior to insertion in the steam generator tube. When the hydraulic expansion tool is pressurized, the bladders act directly against the inside diameter of the sleeve causing expansion of the sleeve.
The sleeve is expanded into contact with the tube and produces a minimal tube diametrical expansion at the joint location. The process utilizes either a pressure control or volume control system. The pressure system is controlled by a pre-set expansion pressure which is determined based upon site specific sleeve-tube material strength. The volume system is controlled by injecting a fixed volume of water into the bladders upon sleeve to tube contact.
4.5.6 Sleeve Weldino Fnninment The welding equipment which performs the gas tungsten arc sleeve to tube welds is comprised of two major components; the sleeve weld tool (Figure 4-12) and the weld power supply (Figure 4-13). The weld tool contains a copper wand which is used to hold and conduct the power to the tungsten electrode. The weld tool includes a stainless steel sheath to prevent damage during insertion of the weld tool into the
. steam generator tube and sleeve. Passages within the weld tool are provided for the shield gas to reach the weld tool electrode.'
The weld tool is rotated by the tool driver with the weld tool at the proper elevation.
A hard stop on the weld tool shaft ensures proper vertical positioning. The electric current and shielding gas are delivered to the weld tool electrode by connections at the bottom end of the weld tool shaft.
Report No. CEN-629-NP, Revision 03-NP Page 4-7
ABB Combustion Engineering Nuclear Operations The welding power source is pre-programmed to supply argon shielding gas and pulsating D.C. current in four distinct power output levels. Current output from level one is set to initiate the arc and form the weld " puddle". As weld heat buildup increases, current output from each subsequent level is decreased in order to maintain consistent weld penetration and height. Weld essential variables, including current, voltage, tool rotation speed and coverage gas flow are outlined in the applicable Weld Procedure Specification (WPS).
4.5.7 Nondectructive Framination Three types of nondestnictive examination equipment are utilized during the sleeving process; ultrasonic test (UT) equipment (Figure 4-14), visual test (VT) equipment and (Figure 4-15) eddy current test (ET) equipment.
Ultrasonic testing using an immersion technique with demineralized water as a couplant is used to inspect the tube to sleeve weld. A one-quarter inch diameter focusing transducer is positioned in the weld area and rotated by the tool driver to scan the weld. A digital imaging system is used to acquire and store the inspection data.
Visual inspection of the steam generator tube to sleeve weld is accomplished with the use of a boroscope or micro camera system delivered and rotated by the tool driver.
Inspection data is stored on video tape.
The ET inspection is performed using the most recently developed eddy current probes and techniques for sleeving inspection. The eddy current probe presently being used is the new advanced + point rotating probe. Future probe designs may be used after suitable qualification program has been performed. The ET guide tube and conduit is mounted on the manipulator arm, which is used to position tie probe.
4.5.8 Postwgid Heat Trotment Fnninment When it is considered to be of benefit, a postweld heat treatment of the sleeve weld is included as part of the sleeve installation process. The postweld heat treatment is performed with a resistarse heater designed to heat the weld, the weld heat affected zom gnd the pressure boundary portion of the tube (either 1/2 inch above, for an FDTS sleeve. or upper TS sleeve weld, or 1/2 inch below for a lower TS sleeve weld), 'The tempersture control is acomplished by thermocouple measurement of the temperature of the resistance heater. The thermocouple readings are input to a controller which initiates the heat treatment process and maintains the heater at a pre-set temperature. The pre-set heater temperature, up to 1900*F, has been qualified to heat the weld joint and adjacent tube wall to a temperature within the range of 1300 to 1425'F for 3 to 5 minutes. Other temperature monitoring methods may be applied to the heaters if they are qualified to provide equal or better control of the Report No. CEN-629-NP, Revision 03-NP Page 4-8
. ABB Combustion Engineering NucleOr Operations
}
temperature of the weld and adjacent tube wall. The PWHT tool is shown in Figure i
4-16.
4.5.9 Kleeve Rolling /Tanering Fnninment i
The sleeve rolling equipment is used to expand the FDTS sleeve (with rolled lower joint) into contact with the steam generator tube within the tubesheet, forming a strong leaktight joint. The sleeve tapering equipment is used to provide tight contact l
with the steam generator tube end for welding the FDTS sleeve lower joint. The rolling tool is mounted on the manipulator and positioned within the sleeve by a hard stop on the rolling tool shaft which seats against the tube end. The rolling tool includes a dovetail attachment for quick mounting on the manipulator. The rolling j
equipment, shown in Figure 4-17, may be used in both the central tubesheet region l
and the periphery region.
l The rolling equipment consists of the air motor, tube expander and a torque l
calibration unit. The torque readout and settings of the rolling tool are verified on the torque calibration unit prior to rolling of the sleeves. The rolling tool is located by a hardstop on the tool shaft. The hardstop positions the upper end of the tube.
expander within the portion of the sleeve which was hydraulically expanded during sleeve installation. The approximately two and one half inch long roll is located at the nickel and metal oxide bands on the lower end of the FDTS sleeve. The-sleeve is egnawi~i to a torque which has been demonstrated by testing to provide a leaktight joint. A record of the rolling tool torque is made for further evaluation of the rolling process on the individual sleeves. A rolled joint which fails to meet the acceptance criteria may be rerolled.
4.5.10 Perinheral % eve Twdlatinn Fnninnwnt The tooling for installation of peripheral sleeves utilizes the proven technology of the tooling used for installation of straight FDTS sleeves.
In addition to the equipment listed in this section, the installation of peripheral sleeves employs a sleeve straightening tool, shown in Figure 4-18. The straightening
. tool consists of multiple rolls which straighten and deliver the sleeve into the tube.
l The curved sleeve, located on the installation expansion tool, is ' installed with the straightening tool until the stops on the installation tool limit sleeve insertion. The expansion is formed in the area of the upper weld joint and lower rolled joint, which acts to hold the sleeve in place for subsequent operations.
In' designing this sleeve, particular attention was paid to minimizing the residual stresses in the straightened section of the sleeve. This involved specific requirements for the fabrication process, including heat treatment, sleeve radius of curvature and the design of the installation tooling. This' process produces a sleeve with post installation residual stress in the straightened portion of no more than [
]. Once Report No; CEN-629-NP, Revision 03-NP Page 4-9
1 ABB Combustion Engineering Nuclear Operations the sleeve is straightened and installed in the steam generator tube, the remaining
)
operations are identical to those for the straight FDTS sleeves.
4.6 ALARA CONSIDERATIONS The steam generator repair operation is designed to minimize personnel exposure during installation of sleeves. The manipulator is installed from the manway without l
entering the steam generator. It is operated remotely from a control station outside the containment building. The positioning accuracy of the manipulator is such that it can be remotely positioned with a high degree of repeatable accuracy.
The tooling delivery equipment is designed such that the dovetail fitting quickly attaches to the manipulator. The tool driver is designed to quickly engage the individual sleeving tools. The tools are simple in design and all sleeving operations are performed remotely using tools held by the manipulator.
Each tool can be changed at the manway in 10-15 seconds. A process step is performed on several sleeve locations rather than performing each process step on the location before proceeding to the next location. This reduces the number of tool changes which are required. Spare tools are provided so that tool repair at the manway is not required.
If tool repair is necessary, the tool is removed and sleeve operations continue using a spare tool. The tool may or may not be repaired during the outage but repair is performed in an area which does not result in significant radiation exposure to personnel.
Air, water and electrical supply lines for the tooling are designed and maintained so that they do not become entangled during operation.
This minimizes personnel exposure outside the steam generator. All equipment is operated from outside the containment building. The welding power supply is stationed about a hundred feet from the steam generator in a low radiation area.
Lead lined manway shield doors, both primary side and secondary (ventilation) side, are also employed to reduce radiation exposure.
4.7 REFERENCES
FOR SECTION 4.0 4.7.1 Alloy 690 for Steam Generator Tubine Apolications, EPRI Report NP-6997, October 1990.
4.7.2 Sedricks, A. J., Schultz, J. W., and Cordovi, M. A., "Inconel Alloy 690 - A New Corrosion Resistant Material", Japan Society of Corrosion Engineering,28,2 (1979).
4.7.3 Airey, G.
P., ' Optimization of Metallurgical Variables to Improve the Stress Corrosion Resistance of Inconel 600", Electric Power Research Institute Research Program RP1708-1 (1982).
Report No. CEN-629-NP, Revision 03-NP Page 4-10 l
l ABB Combustion Engineering Nuclear Operations l
4.7.4 Airey, G. P., Vaia, A. R., and Aspden, R. G., "A Stress Corrosion Cracking j
Evaluation of Inconel 690 for Steam Generator Tubing Applications", Nuctenr j
Technnlogy,15, (November, 1981) 436.
l j
4.7.5
- Hunt, E.S.
and
- Gorman, J. A.,
Snecifications for In-Situ Stress Relief of PWR Reenm Generator Tube U-bends and Roll Trancitinn, EPRI Report NP-4364-LD, Electric Power Research Institute, Palo Alto, CA, December 1985.
]
4.7.6 Krupowicz, J. J., Scott, D. B., and Fink, G. C., " Corrosion Performance of
{
Alternate Steam Generator Materials and Designs Vol. 2: Post Test Examinations of j
a Seawater Faulted Alternative Materials Model Stea n Generator," EPRI-NP-3044, July 1983.
4.7.7 G. Santarini et al, Recent Corrosion Results - Alloy 690, EPRI Alloy 690 Workshop, New Orleans, LA, April 12-14, 1989.
l
]
i i
I l
1 i
i i
Report No. CEN-629-NP, Revision 03-NP Page 4-11
I ABB Combustion Engineering Nuclear Operations I
0.743" Diameter I
n s
l l
l 0.037" Minimum
/
30" l
i t
0.820" Diameter Prior to Welding i
i
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l 1r h
FIGURE 4-1 FULL DEPTH TUBESHEET SLEEVE FOR A WELDED LOWER JOINT Report No. CEN-629-NP, Revision 03-NP Page 4-12
1 ABB Combustion Engineering Nuclear Operations 0.743" Diameter a
I i
l 1
1 1
0.037" Minimum i
/
30" 1
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0.5 inch 1
0.002"/0.004" Microlok Coating 5
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T, 0.002"/0.004" Nickel d
d v
0.5 inch v
FIGURE 4-2 FULL DEPTH TUBESHEET SLEEVE FOR A ROLLED LOWER JOINT 1
I l
i Report No. CEN-629-NP, Revision 03-NP Page 4-13
ABB Combustion Engineering Nuclear Operations i
STRAIGHT SECTION
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FIGURE 4-3 FULL DEPTH TUBESHEET PERIPHERAL SLEEYE Report No. CEN-629-NP, Revision 03-NP Page 4-14
ABB Combustion Engineering Nuclear Operations e
0.743" Diameter
.1 1
a 1
i 4
1
-+
+-- 0.037" Minimum j
l l
9.00"
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1 v
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FIGURE 4-4 TUBE SUPPORT SLEEVE Report No. CEN-629-NP, Revision 03-NP Page 4-15
ABB Combustion Engineering Nuclear Operations
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TUBE i
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SLEEVE H
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WELDED JOINT FIGURE 4-5 FDTS SLFEVE INSTALLATION WITH A WELDED LOWER JOINT Report No. CEN-629-NP, Revision 03-NP Page 4-16
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ABB Combustion Engineering Nuclear Operations
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TUBE i
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y SLEEVE ff d
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i ROLLED JOINT FIGURE 4-6 FDTS SLEEVE INSTALLATION WITH A ROLLED LOWER JOINT Report No. CEN-629-NP, Revision 03-NP Page 4-17
ABB Combustion Engineering Nuclear Operations 4
TUBE gj f
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TUBE SUPPORT SIREVE INSTAILATION j
Report No. CEN-629-NP, Revision 03-NP Page 4-18
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im FIGURE 4-12 ST FFVE WFIBING TOOL l
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l FIGURE 4-13 ST FFVE WEI DING POWER SUPPLY UNIT 4-24
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FIGURE 4-14 ULTRASONIC TEST TOOL 1
4-25 l
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I FIGURE 4-15 VISUAL TEST EOUIPMENT 4-26
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FIGURE 4-16 POSTWFT D HEAT TREAT TOOL 4-27
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manese b
l FIGURE 4-17 SI FFVE ROILING EO_UIPMENT l
4-28 1
m FIGURE 4-18 PERIPIIERAL Si i:I:VE INSTAT T.ATION TOOL 4-29
. - - - - - - - -... - -. - - -.. -. - - ~ -. -. -. -
lL ABB Combustion Engineering Nuclear Operations
]r-5.0~
SLEEVE EXAMINATION PROGRAM i
l' During'the installation process, the sleeves are examined using a combination of L
visual testing (VT), ultrasonic testing (UT) and eddy current testing (ET) at different j
i
_ stages of the installation process. The general process is described in the flow charts presented in Figures 5-1 and 5-2, which are described below.
After the description of the inspection process, the individual inspection methods will
}
be described in additional detail.
[
After completion of the brush cleaning step, the first inspection is a VT process on
{
tubes to be sleeved to confirm adequate cleaning to proceed with the welding process.
j-Parent tube cleanliness has been identified as a critical feature of the overall welding process. A VT after cleaning is performed with a miniature remote camera inserted l
into the tube up to the elevation where the welding will be performed. The VT l
j.
inspectors are. trained using images of examples of acceptable and inadequate cleaning. In simplest terms, the cleanliness requirement is the presence of " bright, 2
shiny metal" in the region of the tube where welding will take place. If adequate l
cleaning is not confirmed by the remote VT, then the cleaning process is repeated j
until a suitable cleanliness is achieved. The extent of this inspection program is presently 100% of tubes to be sleeved.
At such time that process control is demonstrated to assure cleaning efficiency, a sampling program may be used.
i Upon confinnation of cleaning, the sleeve is inserted, expanded and welded. The i
next inspection is performed on the ATS weld by UT to confirm a leaktight bond has been achieved by the welding process. The weld height is not measured by the UT j
method, but rather is controlled by the welding process qualification. A confirmation i
of 360 degrees of weld bond is the acceptance criteria for the UT inspection. If a lack of fusion (LOF) through the weld height is detected, then the sleeve may be identified for rewelding or plugged. After a reweld, the UT is repeated to confirm a l'
leaktight weld. An acceptable UT result is required for any ATS weld left in service.
Prior to the UT inspection, an optional VT-1 inspection of the ATS weld may be performed, but is not required. The VT-1, as defined in ASME Section XI, is j
suitable for detection of incomplete welds, blow holes and weld splattered geometric irregularities in the weld. Experience has shown that the UT and ET inspections are capable of detecting these conditions, so the VT-1 is primarily useful to help resolve uncertainties in surface conditions detected by either the UT or FT inspections. If a VT-1 inspection is performed and a blow hole or other potentially deleterious j
condition (with the exception of an. incomplete weld) is detected, then a
[
noncomformance repon (NCR) must be generated. Blow holes identified as within
{
the pressure boundary portion of the weld must be repaired. Blow holes not within the pressure boundary portion of the weld are identified for additional evaluation by the ET and UT inspections.
l I
l Report No. CEN-629-NP, Revision 03-NP Page 5-1 l
l-JABB Combustion Engineering Nucl:ar Operations A VT-1 inspection of 100 % of all FDTS sleeve lower welds is required.
l l
i The final inspection is performed on all installed sleeves using the ET method with a
+ point probe. If postweld heat treatment is performed, this inspection must be performed after the heat treatment due to the possibility of additional signals from l
permeability variations caused by the heat treatment process. The entire length of the l
pressure boundary, including the pressure boundary portion of the parent tube behind l
l the sleeve is inspected with the ET method. The details of the ET inspection are j
L described in Section 5.2 and Figure 5-2 with the associated definitions in Table 5-1.
The sleeve to tube weld joints are qualified by process control as described in Section
- 9.0. Checks are made to ensure that the welds meet these design requirements. The welding current and voltage are recorded as the weld tool rotates inside the sleeve, i
The weld traces are examined after the welding sequence has been completed in order j
to verify that certain essential parameters defined in the qualification procedure are j
met.
These descriptions of inspection techniques and tooling represent the current state-of-j the-art practices. As new technology becomes available, advanced techniques may be j
substituted after a suitable qualification program has demonstrated equivalent or superior performance.
5.1 ULTRASONIC INSPECTION 5.1.1 hmmarv nad Cand=ians An ultrasonic inspection is performed on each sleeve to tube ATS weld to confirm a leaktight fusion. The test is performed using an ultrasonic crystal with a resonant frequency of [ ] MHz. Physical constmetion of the probe will reduce the effective output frequency to [
] MHz, typically. Actual output frequency is documented in the transducer certification package required by procedure. The mechanical drive device performs a scan of the weld in 2 degree increments around 360 degrees with axial step increments of [
] inches; the scan path extends from above the weld so that the sleeve backwall is detected to below the weld until the backwall of the sleeve is detected. The inspection is demonstrated to detect a milled notch representing a weld lack of fusion (LOF) region of [
] inch or greater. The ultrasonic signal is digitized and stored in order to provide a permanent record of the individual A scans (lower presentation on Figure 5-3), which are used to display plan view C scans L (upper presentation on Figure 5-3) of the weld as well as cross sectional views in the axial direction (B' scans) and cross sectional views (B scans). For each individual l
sleeve inspection, a calibration confirmation is available by monitoring the response l
to the sleeve back wall either above or below the weld zone.
I-Report No. CEN-629-NP, Revision 03-NP Page 5-2 l
l
- - - - -. ~
i ABB Combustion Engineering Nuclect Operations j
5.1.2 Ultrunnic Evahintion i
The basis of the UT inspection is the detection of a reflective surface at the sleeve to tube interface to detect a condition indicative of a lack of fusion.
Sound is I
transmitted from the sleeve inner surface through the weld to the tube outer surface.
Although the reflection from the tube outer surface is typically discernible in the i
recorded data, this is a sufficient, but' not necessary indication of fusion. Geometric l
J distortions in the weld region may preclude detection of this tube back wall as a consistent indicator of weld fusion.
I In the data acquisition phase, a C scan is displayed for the operator with a [
i
] for monitoring reflections from the i
sleeve-tube interface. During analysis, both circumferential and axial cross sectional views of the ultrasonic reflectors are reviewed for evaluation of each weld. Detection i
of a [
] reflection is an indication of a complete weld.
In the absence of this signal, axial and circumferential cross section (B and B' scans) data reviews are conducted. Locally, reflectors are compared to 20% of the sleeve l
1 backwall signal amplitude for determination of a local LOF. Using the B' scan axial cross section, a LOF condition through the weld height is discernible. Using a combination of laboratory samples and removed tubes (Prairie Island, February, j
j 1996), unbonds as narrow as 10 degrees are detectable using this B scan analysis i
technique, as reported in References 5.4.1 and 5.4.2. Sample outputs from the UT i
results for an acceptable weld and unacceptable LOF condition are provided in s
l Figures 5-3 (acceptable) and 5-4 (rejectable).
t i
5.1.3 Test Faninment 3
The test equipment for the ultrasonic inspection comprises the following:
- 1. IntraSpect Ultrasonic Imaging System
- 2. Sleeve Weld UT Inspection Probe,15 MHz, 0.250" diameter crystal, sized for sleeve ID, as shown in Figure 4-14 l
- 3. Couplant supply system, integral with the probe and driver system
- 4. Position device for rotational and translational motion, include encoder feedback for each axis
- 5. Calibration standard with machined notches for initial set up, as depicted in Figure 5-5 Report No. CEN-629-NP, Revision 03-NP Page 5-3
i ABB Combustion Engineering Nuclear Operations
-5.2 EDDY CURRENT INSPECTION
.l 5.2.1 mckerouna For the initial installation of sleeves, each sleeve will be inspected for a baseline and for acceptance. Over the years, the eddy current probe technology has evolved with l
ever increasing sensitivity.in the probe response. Early sleeving programs used a cross wound bobbin coil design, which was later replaced by the I coil design and ultimately by the plus point probe design. The current practice uses the plus point l
probe design with the option of adopting future probe designs after suitable l
qualification demonstration has been performed. The description below discusses the most recent plus point probe. design, which was extensively qualified for sleeve l
inspections in a program that exceeded the requirements of the EPRI Steam Generator
]
Inspection Guidelines, Appendix H in effect at this writing, as described in reference 5.4.3. This qualification used a detection threshold of 40% degradation of the sleeve wall thickness rather than the 60% allowed by Appendix H to add conservatism to the l
f process.
The ET method is used to inspect the entire sleeve region pressure boundary which has four distinct regions:
- 1. the sleeve between the upper weld and lower joint (either roll or weld, depending on sleeve type)
- 2. the pressure boundary region of the steam generator tube behind the sleeve J
- 3. the steam generator tube below the lower rolled joint for an FDTS sleeve j
- 4. the unsleeved portion of the steam generator tube The first three regions are the subject of this discussion, the fourth region is handled as part of the normal tube inspection using the prevailing methods. If postweld heat i
treating is performed on the weld zone, the ET inspection is performed after the heat l
treatment.
l e
5.2.2 Plus Point Probe Onalificatinn Ktudy i
The plus point ET technique was extensively qualified for each of the regions '
identified above using laboratory samples with EDM notches and laboratory produced
-weld imperfections. The details of the inspection samples and results for the weld zone indications are provided in references 5.4.1 and 5.4.2 and the Appendix H i
. qualification report is provided in reference 5.4.3. The Appendix H qualification report provides the details for both the acquisition (ACTS) and analysis (ANTS) of r
the inspection data'.
Report No. CEN-629-NP, Revision 03-NP Page 5-4
ABB Combustion Engineering Nuclear Operations Site specific analysis guidelines have been developed and analysts are trained and tested on the specifics of the technique. In summary, the plus point technique was; demonstrated to be able to detect relevant flaw mechanisms 40% throughwall and greater in each of the regions identified above.
Particular attention was paid to the ATS weld region of the sleeve. The detailed process for the initial installation inspection is shown in the flow chart in Figure 5-2 with the companion list of acronyms in Table 5-1.
For the subsequent inservice j
inspections, reviews of previous inspection results may be used in lieu of the VT and i
UT reviews mentioned in the flow chart. Either the standard + point probe or the
)
magnetically biased style may be used for the inspection. Experience has shown that i
one of the most common interfering signal sources in the weld region is caused by
]
local permeability variations, which are greatly reduced by the partial magnetic j
saturation provided by the magnetically biased probe.
l The ET indications are separated into two broad categories, surface and subsurface.
f Surface indications are caused by minor weld sag which produces a signal classified l
as GEO for geometric. Iacal irregularities in the weld surface are classified as weld surface indications (WSI). In extreme cases, the WSI source could be a blow hole in i
the weld. Additional VT reviews are used to evaluate surface related indications prior to acceptance. With the aid of the VT data, WSI signals are resolved as blow l
holes outside or within the pressure boundary portion of the weld (BHA or BHB) or nondeleterious surface irregularities (WSS). If no surface condition is observed, then the signal is considered as a subsurface weld zone indication (WZI) and evaluated j
accordingly.
For blow holes, the location relative to the pressure boundary is determined using a combination of the VT and UT results. Accordingly, the BHA i
(blow hole outside pressure boundary portion of the weld) condition is acceptable for j
service while the BHB (blow hole within the pressure boundary portion of the weld) is.not.
I i
f The WZI signals may be caused by oxide inclusions in the weld or a partial void l
caused by a gas pocket during the welding process.
Metallographic work, as reported in reference 5.4.1, has shown that these conditions occur at either the upper l
or lower edge of the ATS weld on the sleeve outer surface. The oxide inclusion condition is generally precluded by proper cleaning, which is ierified using VT j
before installing the sleeve. Minor voids may occur in a small percentage of welds even with proper cleaning, but generally are very shallow. No attempt is made to distinguish inclusions from voids, nor is there an attempt to measure depth or circumferential extent for these conditions. The only acceptance criteria is based on the location relative to the pressure boundary with indications outside the pressure l
boundary portion of the weld (WZA) acceptable for service and indications within the l
pressure boundary portion of the weld (WZB) not acceptable for service. The ability j
to determine the true location of indications relative to the pressure boundary portion of the wcid was demonstrated in the Appendix H qualification study and is reported l'
in references 5.4.2 and 5.4.3.
I Report No. CEN-629-NP, Revision 03-NP Page 5-5
_ _. _. _ _ _ _ _ ~. _. _ _ _ _ _ _ _ _. _ _. _
d ABB Combustion Engineering Nuclear Operations i
1 i
Various other anomalous conditions may be reported by the ET analyst that would i
trigger an nonconformance report (NCR) and additional evaluation.
}
The other area of particular interest is the expansion transition zone above the weld.
i Here the parent tube constitutes the pressure boundary. The ability to detect 40%
through wall flaws was demonstrated using EDM notches and is detailed in reference 5.4.3.
1 i
1 5.3 VISUAL INSPECTION i
5.3.1 Summarv and Cancincians j
i There are three visual inspections associated with the sleeving process. The first
]
l inspection is performed after the cleaning process for the weld region. Tubes are l
inspected for cleanliness prior to sleeve installation. The second, optional inspection
~
s is performed after completion of the ATS weld and is conducted as a VT-1 inspection l
per Section XI of the ASME Code. The VT-1 inspection is performed when needed l
to resolve surface indications identified by the ET or UT inspections. The VT-1 inspection is also performed for rewelds. The third inspection is a required VT-1 y
i performed on the FDTS sleeve lower weld.
The VT is performed remotely by means of a miniature CCD camera inserted into the
~
j tube with the results recorded on video tape. Visual aids are provided fer the inspectors for evaluation of cleaning and weld quality.
A training tape with
}
examples of weld irregularities is provided and myiewed by the VT-1 inspectors.
l Conditions of interest include blow holes, pin holes, incomplete welds, splatter and l
pits.
5.3.2 Clannine Inananian
}
]
After the cleaning operation, the parent tube in the region where the weld will be j
made is inspected for adequacy of cleaning. Approximately a two inch long zonc is i
cleaned and inspected. The acceptance criteria is bright, shiny metal to assure that i
j there is no remaining oxide on the tube surface that could affect the weld quality by l
j producing inclusions. This process verification step is identified in the site specific l
]
traveller and is a QC check point required for each tube.
The extent of this j
inspection program is presently 100% of tubes to be sleeved. At such time that l
process control is demonstrated to assure cleaning efficiency, a sampling program j
{
may be used.
5.3.3 Weld Examination j
i The primary inspection methods for ATS weld and sleeve acceptance are the UT and i
ET methods, as described above. An additional VT-1 inspection of the weld is j
optional, unless required by the site procedure for specific situations, such as repair l
I Report No. CEN-629-NP, Revision 03-NP Page 5-6 i
ABB Combustion Engineering Nuclear Operations welds. The VT-1 inspection is also used as a supplemental technique to aid in the analysis of surface conditions reported in either the UT or ET results. The primary inspection method for a FDTS lower weld is a VT-1 inspection of the weld. This is a required inspection for this type of weld.
Prior to the inspection, the camera system is checked by viewing a 1/32" black line on an 18% neutral gray card. Also, a sleeve sample with a 0.020" diameter through hole is used to scale the image. The CCD camera with right angle viewing mirror is inserted into the sleeve, positioned at the weld and rotated 360. The VT-1 results are recorded on video tape for review by a process engineer and for permanent storage.
5.4 REFERENCES
FOR SECTION 5.0 t
5.4.1 CEN-628-P, Rev 01-P, " Verification of the Structural Integrity of the ABB CENO Steam Generator Welded Sleeve," March 1996.
5.4.2 ABB CENO %-3-9038T, Rev 01, " POD Assessment for NDE of Sleeves," June 14, 1996.
f 5.4.3
%-OSW-003, "EPRI Steam Generator Examination Guidelines Appendix H Qualification for Eddy Current Plus-Point Probe Examination of ABB CENO Welded Sleeves," April 27,1996.
f i
Report No. CEN-629-NP, Revision 03-NP Page 5-7
ABB Combustion Engineering Nuclear Operations I
TABLE 5-1 l
ACRONYMS USED IN ET ANALYSIS i
i 1
BHA: Blow Hole Outside Pressure Boundary 1
BHB: Blow Hole Within Pressure Boundary l
GEO: Geometric signal
[
i LOF: Lack Of Fusion NCR: NonConformance Report NDD: No Detectable Degradation i
PID: Positive ID retest I
l RMB: Retest with Magnetically Biased probe i
UT: Ultrasonic Test VT-1: Visual Test, Type 1 per ASME Code,Section XI f
l i
VT: Visual Test l
WEE: Weld at Edge of Expansion l
i WOE: Weld Outside Expansion region l
WSI: Weld Surface Indication l
l WSS: Weld Surface Signal j
l WZA: Weld Zone indication Outside Pressure Boundary WZL: Weld Zone indication Within Pressure Boundary i
WZI: Weld Zone Indication-subsurface or indeterminant
{
i i
Report No. CEN-629-NP, Revision 03-NP Page 5-8 l
2 b
i ABB Combustion Eingineering Nuclear Operations I
l REJECT FLOW CHART NO.1 SIN
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I f
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(PLUG) i pLETg
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BLOW HOLE ACCEPT c
p ACCEPT i
UT REJECT ACCEPT I
ECT p
SEE CHART No. 2 l
FIGURE 5-1 NDE PROCESS FLOW CHART Report No CEN-629-NP, Revision 03-NP Page 5-9
4 f
i ABB Combustion Engineering Nuclear Operations i
l d
h WB FLOW CHART NO,2
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(PLUG) i gummu 2
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FIGURE 5-2 i
i ET PROCESS FLOW CHART i
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)
Report No. CEN-629-NP, Revision 03-NP Page 5-10 l
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~i ABB Combustion Engineering Nuclear Operations r
6.0 SLEEVE-TUBE CORROSION TEST PROGRAM I
i i
j ABB-CE has conducted a number of bench and autoclave tests to evaluate the corrosion resistance of the welded sleeve joint. Of particular interest is the effect of the mechanical expansion / weld residual stresses and the condition of the weld and weld heat affected zone. Tests have been performed on welded joints with and
[
7' without a'postweld heat treatment. An outline of these tests is shown in Table 6-1.
As shown in the table some tests have been conducted with mill annealed Alloy 690 j
' sleeves due primarily to material availability. For the environments being tested l
1 ABB-CE considers these to be conservative tests in that thermally treated material has
}
been shown to perform as well or better than mill annealed material.
I' 6.1
SUMMARY
AND CONCLUSIONS i
- Tests conducted indicate that the welded sleeve-tube joint performs well in corrosion j
tests designed to simulate typical fault and normal conditions on an accelerated basis, i
Generally corrosion under anticipated service conditions is indicated oldy for the steam generator tube and not for the sleeve or weld metal.
6.2 TEST DESCRIPTION AND RESULTS i
6.2.1 Primuv We Tests i
Certaia metallurgical conditions of Ni-Cr-Fe Alloy 600 have experienced stress corrosion cracking in prunary sid6 water. To ensure the viability of the welded j
sleeve design, ABB-CE has pursued two approaches in evaluating its performance i
under these environmental conditions, i
The first involved testing of sleeve-tube assemblies under accelerated pure water conditions. Although such tests have provided some measure of the performance of as-welded sleeved tubes, it was recognized that a long term estimate of the life of sleeved tubes could not readily be made from such data. An alternate approach was thereby undertaken in which the stress on the joint is utilized to estimate the life of the sleeved tube. In order to assess the sleeved tube's stress state, corrosion testing is performed in an environment (10% NaOH) known to stress corrosion crack Alloy 600 in a relatively short period of time.
This data is then combined with the local i
operational stresses, as determined by finite element analysis, to calculate the total stress on the sleeved tube. By the use of the relationship for primary side stress corrosion cracking presented in Reference 6.3.1 and actual plant experience, an estimate of the installed sleeve life can be made.
i-Report No. CEN429-NP, Revision 03-NP Page 6-1
ABB Combustion Engineering Nuclear Operations TABLE 6-1 STEAM GENERATOR TUBE SI FFVE CORROSION TESTS TEST MATERIAL ENVIRONMENT a.
7
?
- (1)
Specially Prepared Mill Annealed and Ringhals 2 Tubing (2)
Thermally Treated (3)
Tube Material Supplied by EPRI (B&W Experimental Heat No.%834, Lot 2)
(4)
Tube Material Supplied by Swedish State Power Board (5)
AW - As Welded PWHT - Postweld Heat Treated Report No. CEN-629-NP, Revision 03-NP Page 6-2
r ABB Combustien Engineering Nuclear Operations 6.2.1.1 Pure Water Stress Cor-osion Craino Tests ABB-CE has conducted autoclave tests in which as-welded sleeve-tube assemblies (Figure 6-1) are exposed to a pure water environment at 680*F. The steam generator tubing used for this test included material removed from the original Ringhals 2
, team generators and specially processed Alloy 600 tubing particularly susceptible to primary water stress corrosion cracking (PWSCC). [
).
The joints in this test also represent a worse case condition with respect to the mechanical expansion used to provide fit up for the welded joint. The process as qualified requires a 0.005 inch maximum diametrical expansion of the tube O.D.
Specimens tested have a tube O.D. expansion on the order of 0.025 inches.
These specimens accumulated up to [
] hours on the specially prepared mill annealed tubing and up to [
] hours on the Ringhals 2 tubing without cracking.
6.2.1.2 Above the Tubesheet (ATS) Weld Capsule Tests In many potential applications for tube sleeves, a high enough total stress in the parent Alloy 600 tube may lead to stress corrosion cracking of the tube if appropriate metallurgical conditions exist. In order to confirm the applicability of the welded sleeve to units with highly susceptible tubing ABB-CE developed a postweld heat-treatment process to lower the total stress in the parent tube after sleeve installation.
In order to qualify this process, actual sleeve samples were exposed on the inside surface to an accelerated stress corrosion cracking (SCC) environment. Based upon the time to cracking of the specimens, a sleeve life could be determined (Reference 6.3.2).
The specimen configuration is shown in Figure 6-2. The tube material used included a special heat (B&W No. %834, Lot 2) prepared for an EPRI ccntract investigating primary side stress corrosion cracking. A 10% NaOH solution at 660'F was used in
- the test assemblies. After welding and heat treating, the sleeve-tube assemblies had a Nickel 200 plug welded to the bottom. They were then filled with solution, capped with a Swagelock fitting, weighed.and placed in a furnace. The sleeved samples were weighed and visually inspected on a daily basis to determine whether through-wall cracking had occurred. C-ring specimens for stress indexing were included in a Nickel 200 capsule in the same furnace. When a cracked sleeve-tube assembly was discovered, C-rings with a comparable amount of exposure time were removed and examined for cracking.
Report No. CEN-629-NP, Revision 03-NP Page 6-3
i ABB Combustion Engineering Nuclear Operations -
)
l The results of these tests are shown in Table 6-2.
j l
L TABLE 6-2 1
Ph:wTh r=:=' SCC Teate l
hmnie condisinn Failure No.
Failure Time (Hrd l
}
l l-Comparing the time to cracking of the sleeve-tube capsule tests with the C-ring L
cracking and adjusting for the applied axial stress in the capsule, it was estimated that i
the as-welded samples had a residual stress of [ ] ksi and the PWHT samples had a i
. residual stress of much less than [ ] ksi.
j l
Using data developed from the test program and finite element analysis of the local stresses, the life of the sleeve relative to the rest of the steam generator could be i
estimated using the relationship described in Reference 6.3.2.
l f
l 6.2.1.3 TSP Sleeve Weld Capsule Tests
. In addition to the previously described tests for heat treated ATS sleeve welds, ABB-l CE has conducted tests on specimens in the TSP configurations. These tests also j
employed specimens exposed to 10% NaOH on the inside surface.
The tube
- material, which is highly susceptible to PWSCC, was supplied by the Swedish State Power Board.
t L
Baseline roll transition specimens were tested as an index for the caustic capsule l
specimens. These specimens consisted of a tube rolled into a block to the same wall thickness reduction as in production joints, internally pressurized and exposed to the 10% NaOH environment. Failure times for these specimens ranged from [
L l
-] hours with an average of [
] hours.
[
L i
L The TSP specimens, shown in Figure 6-3, were prepared by installing the sleeves in f
I a tube mockup. The tubes in this mockup were fixed at both ends by welding, j
simulating a locked condition above the sleeve installation location. Due to the tube
. lengths available, the distance from the sleeve location to the locked location was j
shorter than in an actual generator. This condition leads to a more conservative test,
[
since there is minimal tube movement allowed due to thermal expansion during the l
welding and heat treatment process. After heat treating, the specimens were removed
(
from the mockup and prepared with a Nickel 200 plug and Swagelock fitting as previously described. No external loads were applied to the specimens during the j
test exposure, only those loads imposed by the differential pressure of 2200 psi.
l r
L i
l Report No. CEN-629-NP, Revision 03-NP Page 6-4 l
g j
a
ABB Combustion Engineering Nuclear Operations c
i i
These tests exhibited times to cracking ranging from [
] hours before experiencing through wall stress corrosion cracking. One sample was exposed for j
[
] hours without through wall cracking. These times are up to [
]
l greater then for the baseline roll transition specimens.
6.2.1.4 ' Summary - Primary Coolant Corrosion Performance The life of the sleeve to steam generator tube joint has been postulated to be a l
function of its resistance to primary water stress corrosion cracking (PWSCC). Field i
experience has shown this to be the case with other types of sleeve designs. As is well documented elsewhere, this resistance will depend on three parameters; the l
initial metallurgical condition of the tube, the stresses in the tube and the temperature of the joint. Barring operational changes, the temperature of the joint will not differ i
from that of the original parent tube nor in general will the metallurgical condition (in some cases, the grain growth and solution annealing that occurs in the weld heat i
affected. zone may actually improve the resistance). Therefore, the single most i
important factor in assessing the life of the sleeve-tube joint from the primary side is
)
the state of stress associated with the installed sleeve.
Corrosion tests described above on as-welded (without postweld heat treatment) steam generator sleeve-tube assemblies have confirmed that the area with the highest susceptibility to stress corrosion cracking is that adjacent to the weld in the parent tube. ABB-CE has performed analyses and tests to evaluate the total stress state at this critical location. Using this data and the relationship described in Reference 6.3.1, an estimate of the sleeve-tube joint stress state can be made.
- The total stress state associated with this area can be defined by its three components; the applied stress, short range or micro residual stress, and long range or reaction stress.
Annlied Streccec - The applied stresses can be calculated using the finite element analysis. of the sleeve-tube weld joint described in Section 8 for the region immediately adjacent to the sleeve-tube weld joint. The combination of pressure and thermal stresses results in a combined compressive stress on the tube I.D. surface.
- The compressive stress is a result of the greater expansion of the sleeve relative to the tube due to its higher thermal expansion coefficient and higher temperature. This produces a bending moment on the I.D. surface and the corresponding compressive stress.
Micro ReciAn=1 Strecces in the Weld HAZ - The HAZ stresses due to the localized weld shrinkage have been determined as described in Section 6.2.1.2. Under these conditions, stresses for the postweld heat treated joint were found to be much less than [
].
Report No. CEN-629-NP, Revision 03-NP Page 6-5
ABB Combustion Engineering Nuclear Operations penceinn Streccec in locked Tubes - When the tubes are locked at support locations due to denting or other mechanisms, the tube undergoes a compressive load during heating (whether by. welding or PWHT) which shortens the tube and causes some column buckling. During cooling the tube is put into tension. The magnitude of the resulting stresses depends on the amou
- of deformation which is a function of the temperature, length of the heated zone, tra length of the span, the time the tube is heated, and the number of welds in any one tulu In the case of the as-welded sleeve-tube joint, these stresses are on the order of [
). For a tube on which a postweld heat treatment has been performed, the reaction stresses have been determined as desc:ibed in Section 9. In the case of a tube with a single RTZ sleeve in tubes locked at the first support plate, values of [
]
could be expected. A tube with a RTZ and two TS sleeves instalid could result in stresses as high as [
] in some sections of the tube.
Sleeve Joint Life Determination - Once the stress state for the specific plant conditions has been established, comparison with a known failure is made to determine the expected life of the joint. In conjunction with failure times for the specific plant or those of like characteristics, the life of the sleeve can be estimated from the Arrhenius relation described in Reference 6.3.1 Ls=br x(S /Srf 3
where:
Ls = Life of the sleevedjoint Lr = Life of the original tube S = Stress in the sleeve-tube joint 3
Sr = Stress in the original tube failure n = Empirically determined exponent (conservative value =4) 6.2.2 secandary Side Tests 6.2.2.1 Modified Huey Tests The modified Huey test has been used to evaluate the extent of grain boundary carbide precipitation / chromium depletion (sensitization) that may occur in the welds or weld heat affected zone. In some environments Ni-Cr-Fe alloys are susceptible to intergranular attack or intergranular stress corrosion cracking if a sensitized microstructure is present.
Sleeve-tube weld joint sections were cut into 90* segments and subjected to the modified Huey Tests. The conventional Huey Test (ASTM A262-81, Practice C, Nitric. Acid Test for Detecting Susceptibility to Intergranular Attack in Austenitic Stainless Steels) entails boiling specimens in 65% nitric acid solution for five-48 hour Report No. CEN-629-NP, Revision 03-NP Page 6-6
ABB Combustion Engineering Nuclear Operations f
cycles. The modified Huey Test differs in that it uses 25% nitric acid for one-48 l
hour cycle.
Normally, specimens are weighed before and after testing, and the percent weight loss is used to evaluate the susceptibility to intergranular attack. The sleeve samples, however, contained three separate metallurgical structures, i.e., base metal, heat i
Because of this, weight changes could not
(
differentiate weight loss with respect to structure.
Hence, the specimens were l
l examined using a metallurgical microscope and the extent of attack determined l
optically.
l 4
I l
L
?
t i
I L
t 6.2.2.2 Capsule Tests Capsule tests have been employed to evaluate the corrosion resistance of the steam l
[
generator sleeve-tube assembly to a wide range of fault secondary side environments.
L Of particular concern is the crevice area between the sleeve and tube at the weld j
i joint. The specimen for this test consists of a 90* segment of the weld joint. These -
l l
samples were encapsulated in scaled tubes containing the desired environment at l
650 F (343*C).
Specimens were removed at periodic intervals and examined l
l metallographically for corrosion attack. The tests conducted and results are listed in l
t Table 6-2. In summary, [
l
].
The significance of this is that the sleeve wel< ling had benign f
affects on the materials in chlorides, sulfates and resin fines.
l In caustic test environments, the performance of Alloy 690 was superior to that of i
l Alloy 600. [
l i
l The corrosion resistance of Alloy 690 as compared to Alloy 600 has been well l
documented (References 6.3.3 - 6.3.8). In all caustic corrosion tests, the Alloy 690 j
i
- had significantly greater corrosion resistance than Alloy 600.
l t
i i
i i
i
?
l 1
Report No. CEN-629-NP, Revision 03-NP Page 6-7 I
l
j ABB Combustion Engineering Nuclear Operations l
i l
TABLE 6-3 l
l SECONDARY SIDE STEAM GENERATOR TUBE SI REVE CAPSUIR TESTS i
ENVIRONMENT EXPOSURE TIME RESULTS j
y i
I 6.2.2.3 Sodium Hydroxide Fault Autoclave Tests Caustic stress corrosion cracking of Ni-Cr-Fe Alloy 600 has been observed in some l
operating steam generators. The corrosion resistance of Alloy 690 as compared to Alloy 600 has been well documented (References 6.3.3 - 6.3.8).
In all caustic corrosion tests, the Alloy 690 had significantly greater corrosion resistance than Alloy 600. As an example, Berge and Donati (Reference 6.3.6) report that, based on l
crack growth rate, Alloy 690 was an order of magnitude better than Alloy 600 and that thermally treated Alloy 690 was an order of magnitude better than mill annealed Alloy 690.
To simulate the intrusion of a fault caustic condition into the annulus between the l
tube and sleeve, the specimen for this test was designed as shown in Figure 6-5.
[
Holes were drilled into the Alloy 600 tube to allow caustic direct contact with the L
Alloy 690 sleeve.
l l
The test rpecimens were immersed in a nickel container containing 10% NaOH. The nickel container was then immersed in an autoclave which provided the heat and pressure for the test. The autoclave water was heated to 680 F (360 C) at 2708 psi (18.7 MPa) saturation pressure. The hydroxide solution was also heated to 680 F i
i l
Report No. CEN-629-NP, Revision 03-NP Page 6-8
ABB Combustion Engineering Nuclear Operations (360*C). The specimens were internally pressurized to 4900 psi (33.8 MPa) to give a differential of 2200 psi (15.2 MPa). This stress level is significantly greater than normal operating pressures.
Tube sleeve assemblies were tested for up to 3000 hours0.0347 days <br />0.833 hours <br />0.00496 weeks <br />0.00114 months <br /> and removed for eddy current testing and metallographic examination.
Results indicated [
). No attack was evident in any lower sleeve welds. [
3-In addition, the [
].
Welds stressed to twice that expected in service showed no cracking during this exposure.
6.2.2.4 Summary - Secondary Coolant Coitosion Performance Based on this testing the sleeve and sleeve-tube joint have exhibited resistance to various potential secondary side fault environments. It is possible that the local region of the sleeve-tube joint may, depending on the metallurgical condition of the tube and whether postweld heat treatment is applied, be susceptible to one or another secondary fault environment. However, by the very nature of the weld joint in both
-an FDTS and TS sleeve, it must be located outside a region where fault species are capable of concentrating, ie. tube / support intersections. These environments would only exist in the area bridged by the sleeve and as such the structural integrity of the sleeve-tube assembly would not be degraded.
6.3 REFERENCES
FOR SECTION 6.0 6.3.1 Staticticni Analysis of Stamm Generator Tube Degradation, EPRI Report NP-7493, September 1991.
6.3.2. Summary Report, Combustion Engineering Steam Generator Tube Sleeve Residual Stress Evaluation, TR-MCC-153, November 1989.
6.3.3 I. L. W. Wilson and R. G. Aspden, " Caustic Stress Corrosion Cracking of Iron-Nickel-Chromium Alloys." Stress Corrosion Cracking and Hydrogen Embrittlement of Iron Race Alloys, NACE, Houston, Texas, pp 1189-1204, 1977.
Report No. CEN-629-NP, Revision 03-NP Page 6-9
ABB Combustion Engineering Nuclear Operations 6.3.4 A. J. Sedriks, S. Floreen, and A. R. McIlree, "The Effect of Nickel Content on the Stress Corrosion Resistance of Fe-Cr-Ni in an Elevated Temperature Caustic Environment". Corrosion,Vol. 32, No. 4, pp 157-158, April 1976.
6.3.5 F. W. Pement, I. L. W. Wilson and R. G. Aspden, " Stress Corrosion Cracking Studies of High Nickel Austenitic Alloys in Several High Temperature Aqueous Solutions." MaterinkPerformance, Vol.19, pp 43-49, April 1980.
6.3.6 P. Berge and J. R. Donati, " Materials Requirements for Pressurized Water Reactor Steam Generator Tubing." NuclearTechnology, Vol. 55, pp 88-104, October 1981.
6.3.7 G. P. Airey, A. R. Vaia and R. G. Aspden, "A Stress Corrosion Cracking Evaluation of Inconel 690 for Steam. Generator Tubing Applications."
Nuclear Technology, Vol. 55, pp 436-448, November 1981.
6.3.8 J. R. Crum and R. C. Scarberry, " Corrosion Testing of Inconel Alloy 690 for PWR Steam Generators." Journal of Materink for Energy Systems, Vol. 4, No. 3, pp 125-130, December 1982.
j i
J Report No. CEN-629-NP, Revision 03-NP Page 6-10
)
ABB Combustion Engineering Nuclear Operations l
MONEL 400 PLUG C ^
i i
FILLED WITH l
l l
PRESSURIZED
{
P, = P,
-2200 psi j
i i
WATER j
T = 680' F yI TRAPPED GAS REPLACED BY WATER M,
WELD i
P, i
P, c
l l
l l
l WATER l
MONEL 400 WASHER REPLACES GAS I
z,wd j
l tmag ip (i
PIPE CAP i
l MONEL 400 i
l l
FIGURE 6-1 i
PURE WATER CORROSION TEST SPECIMEN Report No. CEN-629-NP, Revision 03-NP Page 6-11
ABB Combustion Engineering Nuclear Oper:tions SWAGELOK FITTING w
--a i
FILLED WITH 10% NaOH r'q' i
Pi =2250 psi N
P, = ATMOSPHERE l
l T
= 6607 P,
P, i l i
l 2 l
is/
h,!
{l i
l i!
l l
FIGURE 6-2 ATS WELD CAPSULE TEST SPECIMEN Report No. CEN-629-NP, Revision 03-NP Page 6-12
ABB Combustion Engineering Nuclear Operations SWAGELOK FITTING 9
--r
-w FILLED WITH 10% NoOH y
q' I
i lL Pi =2250 psi g
l\\
. l P
= ATMOSPHERE m
m il
- t T
= 6,607 WELD P,
P,
- l in
!:E h!
ij i
!s NICKEL 200 PLUG l
t 1
l}
\\i FIGURE 6-3 TSP WELD CAPSULE TEST SPECIMEN Report No. CEN-629-NP, Revision 03-NP Page 6-13
ABB Combustion Engineering Nuclear Operations MONEL 400 PLUG O ^
i 1
FILLED WITH P' =P
-2200 psi PRESSURIZED l
WATER l !
T = 680* F s
TRAPPED GAS i
REPLACED BY t
p,i p
CAUSTIC SOLUTION l
]
10% NaOH SOLUTION SURROUNDS ASSEMBLY 10% NaOH SOLUTION MONEL 400 WASHER REPLACES-GAS
- d d :
3:
l 4:
PIPE CAP MONEL 400 1
FIGURE 6-4 CAUSTIC CORROSION AUTOCLAVE TEST SPECIMEN Report No. CEN-629-NP, Revision 03-NP Page 6-14 i
i
' ABB Combustion Engineering Nuclear Operations l
l 7.0 MECHANICAL TESTS OF SLEEVED STEAM GENERATOR TUBES 7.1
SUMMARY
AND CONCLUSIONS l
l Mechanical tests were performed on mockup steam generator tubes containing sleeves to provide qualified test data describing the basic properties of the completed i
assemblies. These tests determined axial load, collapse, burst and thermal cycling j
capability. A minimum of three tests of each type were performed.
l i
Table 7-1 summarizes the results of the mechanical testing performed on the sleeve-i tube assemblies. The demonstrated load capacity of the assemblies provides an l
adequate safety factor for normal operating and postulated accident conditions. The j
load capability of the upper and lower sleeve-tube joints is sufficient to withstand i
thermally induced stresses in the weld resulting from the temperature differential between the sleeve and the tube and pressure induced stresses resulting from normal i
operating and postulated accident conditions. The burst and collapse pressures of the l
(
sleeve provide a large safety factor over limiting pressure differential. Mechanical j
testing revealed that the installed sleeve will withstand the cyclical loading resulting j
j from power changes in the plant and other transients.
l 7.2 CONDITIONS TESTED The following tests were performed on the cleeve-tube assemblies at. room temperature: axial pull, load cycling, burst and collapse. Loads were applied until l
the point of failure, or in the case of cyclic loading, until the number of cycles exceeded the expected number of cycles for the plant.
7.3 WELDED SLEEVE TEST PARAMETERS AND RESULTS I
7.3.1 Arial Pull Tests l
The ATS weld joints of the FDTS and TS sleeves and the lower weld joint of the FDTS sleeve attaching the sleeve to the steam generator tube are subjected to axial loads resulting from the differential thermal expansion of the sleeve and tube and from pressure differentials during normal operating and postulated accident conditions.
'The tests were conducted by installing sleeve samples representing either the upper or lower end of the sleeve into sections of the steam generator tube. The specimens were loaded in a tensile machine and the sleeve was pushed and pulled relative to the tube until failure occurred. The ATS weld sample yielded at a minimum value of
[
pounds and failure of the ATS joint occurred at a minimum value of [
]
_ pounds. The FDTS lower weld joint yielded at a minimum value of [
] pounds and the joint failed at a minimum value of [
] pounds. This axial load capability is much greater than the calculated maximum load of [
] pounds (See Section i
Report No. CEN-629-NP Revision 03-NP Page 7-1
~
ABB Combustion Engineering Nuclear Operations i
f 8.0). No slippage occurred when the FDTS lower rolled joint samples were pushed to a value of [
] pounds. Samples were then pulled until joint slippage occurred j
at[
] pounds. This axial. load capability is much greater than the calculated j
maximum load of[
] pounds (See Section 8.0).
l Test assemblies for the ATS weld that had been welded for only 330 degrees of the
' sleeve circumference were also axially pull tested. The reduced weld area did not measurably affect the load carrying capability of the upperjoint.
i j
7.3.2 Collanse Testing j
i l
External pressurization tests were performed to demonstrate that the sleeve would not collapse at the maximum secondary side pressure of 1360 psi, which occurs during a l
hydrostatic leak test. The loss of coolant accident (LOCA) produces a maximum
]
external pressure of approximately 1085 psig on the sleeve-tube assembly at 600*F.
i l
The material strength of Alloy 690 is not substantially affected by temperature
]
j changes in these ranges. ' Therefore, the hydrostatic leak test is the limiting case.
I I
The tests for the FDTS and TS sleeves were conducted by pressurizing the annulus
)
l between the sleeve and the tube with demineralized water through a test fitting j
l welded to the tube. Pressurization of the annulus represents a through wall defect in l
the tube typical of a corrosion type penetration. The pressure was increased in 200 l
psig increments until the tube or sleeve became visually oval, the tube bulged l
l appreciably or the welds developed leaks.
j The minimum sleeve wilapse pressure was [
] psig. No dimensional change was
{
noted in the sleeve until the collapse at [
] psig. After failure, the sleeve inside diameter remained dimensionally the same except in the collapse area approximately l
180 degrees from the pressure tap. This condition does not represent a structural l
integrity or safety issue.
All welded FDTS sleeve and TS sleeve welds remained leaktight during the test.
Similarly, both the ATS weld and lower rolled joint for the rolled FDTS sleeve remained leaktight during the the test.
l The minimum collapse pressure of [
] psig represents a safety factor of approximately [
)) over the hydrostatic test pressure which is the largest pressure differential applied externally to the sleeve. This test also confirms the ability of the sleeve to sustain a loss of coolant accident (LOCA) with a maximum postulated external pressure differential of 1085 psi.
i l
Report No. CEN-629-NP, Revision 03-NP Page 7-2 1
ABB Combustion Engineering Nuclear Operations 7.3.3
~ Burst Testing In the event of. a main steam line break (MSLB), the secondary side of the steam generator would be rapidly depressurized, exposing the sleeve-tube assembly to a l
maximum differential pressure of 2560 psi at 667'F applied to the inside diameter of the sleeve. The MSLB differential pressure is less than the hydrostatic test pressure of 3106 psig. The material strength of Alloy 690 is not substantially affected by temperature changes in these ranges. Therefore, the hydrostatic test pressure is the limiting case.
Burst tests were conducted by welding an end cap on the sleeve-tube assembly and pressurizing the inside diameter of the sleeve in increments until sleeve failure.
Sections of tube were welded at the ends of the sample leaving the central eight (8) inches of the sleeve unreinforced by the tube. This represents a more severe case than would result from degradation of the steam generator tube. At [
] psig, the lower tube to sleeve weld on the first test sample began lealdng. The test was stopped and the outside diameter of the sleeve was remeasured. The sleeve was deformed _approximately 0.089 inches from its initial diameter.
A second test specimen leaked at [
] psig. A third test was terminated at [
] psig when a leak developed in the test fixture.
The termination pressure of [
] psig represents a safety factor of greater than [two (2)] times the hydrostatic test pressure and confirms the ability of the sleeve to withstand the main steam or feedwater line break. It also confirms that the sleeve has a factor of safety of approximately [
] compared with primary to secondary l
pressure differential during normal operation.
7.3.4 Inad Cycling Tests Normal operating and transient conditions result in cyclical loading on the steam generator sleeves due to differential thermal expansion of the tube and sleeve as well as the internal pressure differential. Welded joint and rolled joint assemblies were subjected to cyclical load tests to determine the structural capability of the assemblies.
ATS welded joint and FDTS lower. welded joint test assemblies were placed in a tensile machine and mechanically cycled between [
] pounds compression and
-[
] pounds _ compression for [
] cycles at a rate not exceeding four (4) cycles per. minute. The test assemblies were removed from the tensile machine and the welds were liquid penetrant inspected' for cracks. The test assemblies were then reinstalled in the tensile machine and subjected to an additional [
] cycles with the load varying between [
] pounds compression and [
] pounds compression. A final liquid penetrant inspection was performed.
Report No."CEN-629-NP, Revision 03-NP Page 7-3 E
r ABB Combustion Engineering Nuclear Operations The load cycling test was performed on both fully welded test assemblies and test assemblies that had been welded for only 330 degrees of the sleeve circumference.
No cracks were detected in any of the test assembly welds.
Extensive load cycle testing was performed on Roll / Expansion Transition Zone (ETZ) rolled joint test assemblies for a 3/4" tube sleeving program.
The test specimens were placed in a tensile machine and loaded in compression for approximately [
] cycles. Tubesheet loads resulting in a ligament stress of
[
] psi were then applied coincident with the axial loads for an additional [
]
cycles.
IIydrostatic and helium leak testing was conducted to confirm joint effectiveness. All rolled joints were leaktight.
The test program which was performed to support a rolled sleeve-tube joint in 7/8" steam generator tubes built upon the results of the 3/4" test program. The test specimens were placed in a tensile machine and cycled between [ ] pounds and
[
] pounds compression for [
] cycles. The test assemblies were removed from the tensile machine and were hydrostatically leak tested at a pressure of [
]
psi. All rolled joints were leaktight. Additional test specimens were installed in the tensile machine and subjected to a compressive load of [
] pounds. There was no joint slippage at this load.
i
7.4 REFERENCES
FOR SECTION 7.0 7.4.1
" Mechanical Tests of Welded Joints of Steam generator Tubes," TR-MCC-116, April l
1985.
i i
7.4.2
" Steam Generator Weld Integrity Test," TR-ESE-595, April 1984.
7.4.3
" Test Report For The Qualification Of The Roll Transition Zone Sleeve Rolled Joint For Westinghouse "D" Series Steam Generators," TR-ESE-887, April 1991.
j
'7.4.4
" Test Report To Support The Use Of A Rolled Lower Joint In Tubesheet Sleeves For Westinghouse Series "44" And "51" Steam Generators," CEN-631-P, October 1996.
i l
Report No. CEN-629-NP, Revision 03-NP Page 7-4 1
l
ABB Combustion Engineering Nuclear Operations TABLE 7-1 SI.FEVE-TUBE ASSEMBLY MECHANICAL TESTING RERULTS*
COMPONENT AND TEST RESULT RESULT (MAXIMUM)
(MINIMUM) m x i
l i
P f
~
~
- A minimum of three tests of each type were performed.
Report No. CEN-629-NP, Revision 03-NP Page 7-5
F ABB Combustion Engineering Nuclear Operations i
'8.0 STRUCTURAL ANALYSIS OF TUBE / SLEEVE ASSEMBLY This analysis establishes the structural adequacy of the tube / sleeve assembly.
The methodology used is in accordance with the 1995 Edition of the ASME Boiler and f
Pressure Vessel _ Code,Section III.
The work was performed in accordance with 10CFR50 Appendix B and other applicable U.S. Nuclear Regulatory Commission requirements.
8.1
SUMMARY
AND CONCLUSIONS i
Based on the analytical evaluation contained in this section and the mechanical test data contained in Section 7.0, it is concluded that the Full Depth Tube Sheet (FDTS) and the Tube Support (TS) sleeves described in this document, meet all the requirements stipulated in Section 8.0 with substantial additional margins. In performing the analytical evaluation on the tube sleeves, the operating and design conditions for all of the Westinghouse operating plants with 7/8 inch Inconel 600 tubes are considered (Reference 8.2).
t 8.1.1 Iwien mine i
In accordance with ASME Code practice, the design requirements for tubing are covered l
by the specifications for the steam generator " vessel". The appropriate formula for i
calmiatin, the mmunum required tube or sleeve thickness is found in Paragraph NB-l 3324.1, tentative pressum thickness for cylindrical shells (Reference 8.1). _ The following i
calculation uses this formula for the tube sleeve material which is Alloy 690 material with a specified minimum yield of 40.0 ksi.
i l
j Where t = Minimum required wall thickness, in.
I P = Design Tubesheet differential pressure, ksi (max. value for plants, Ref. 8.2) l R = Inside Radius of sleeve, in. (maximum value for plants considered)
S = Design Stress Intensity, S.I. @ 650 F maximum design (per Reference
)
8.15) -
i 8.1.2 Ddmilad Analvsis hmmarv When properly inualled and welded within specified tolerances, the FDTS sleeve and its i
upper weld and lower weld or rolled joint (depending on the tube sleeve design), and the Report No. CEN-629-NP, Revision 03-NP Page 8-1 l
ABB Combustion Engineering Nuclear Operations TS sleeve and its two primary welds possess considerable margin against pullout for all
{
loading which can be postulated from operating, emergency, test, and faulted conditions.
j i
The axial loads in the sleeve are a function of their location within the bundle and of the j
degree of tube / support lock-up. The most severe combinations are determined to be i
[
l 4
] at 100% steady state power for the operating plants in Reference 8.2.
The most severe combinations for the " worst" c== enveinnment of nneratino ennaitions
_j and naramatars in Reference 8.2 are [
]
]
In Section 8.2, a comparison is made between calculated failure modes and test data
}
discussed in Section 7.0 of this report. The agreement between calculated and test data is good. Safety factors are determmed for hypothetical pipe break accidents, and a minimum factor of safety of[
] is determined. The normal operations safety factors of [
] for the
)
sleeve design with upper and lower weld joints and [
] for the sleeve design with upper l
weld joint and lower rolled joint (mimmum for both operating plants and " worst" case l
envelopment) am based on the full power restrained thermal expansion loading. Pusbout at 1
the lower tube / sleeve joint is the critical consideration (see Section 8.4.6).
)
The axial sleeve loads calculated in Section 8.4 are used as boundary conditions and the l
basis for assumptions in the Section 8.6 fstigue evaluations.
I An NRC Regulatory Guide 1.121 evaluation is performed in Section 8.3 to determine when j
a sleeved tube should be plugged. A [
] allowable degradation limit is calculated. l This large allowable degradation is possible hac== the Regulatory Guide is based upon normal operating parameters, such as operating differential pressure, rather than tubesheet design differential pmssure.
I Considerations for susceptibility to flow induced vibration is discussed in Section 8.5.
Based on ABB-CE's expeiiese and test data, it was detennined that a sleeved tube is no more susceptible to vibration than a normal tube.
The minimum required axial length of the weld (see Section 8.8.3) is determined for the 2
upper sleeve to tube weld joint. Fatigue of the upper weld joint is considered in Section 8.6.1. The geometry is shown to meet all ASME Code allowables for the general primary membrane and range of primary and secondary stresses and the fatigue usage factor.
For the sleeve design with upper and lower weld joints, fatigue of the lower weld joint is considered in Section 8.6.3. The geometry is also found to meet all ASME Code allowables like that of the upper weld joint. The tube sleeve geometry considered is shown in Figum 8-
- 1. A tabulation of the results is presented in Table 8-1. The maximum local pnmary stress intensity.is 27.9 ksi at the lower weld joint, ar compared to the allowable of 40.0 ksi. The maximum range of primary plus secondary stress is 46.5 ksi at the upper weld joint, as compamd to the allowable of 80.0 ksi. Table 8-2 contains a tabulation of the lower rolled Report No. CEN-629-NP, Revision 03-NP Page 8-2
_.. _ _.. _ _ _ _. _ _. _.. ~. _ _ _ _. - -. _. _.-- _--._
ll I
L ABB Combustion Engineering Nuclear Operations i
l joint for. the sleeve design with upper weld joint and lower rolled joint versus the i
mechanical test results from Section 7.0.
The sleeve evaluation is performed for a 30 inch FDTS sleeve at the location starting on the pnmary face of the tubesheet. The temperature diffemnces between the sleeve and tube-which affect axial loads occur over a distance between the top of the tube sheet and the j
upper weld. The relative intensity of the developed axial loads is a function of the spring l
constants for the sleeve and tube lengths, the location of the sleeve's lower joint in the l
tubesheet, and the forced displacement between the sleeve and tube.
l t
The evaluation of the TS upper and lower welds shows that the stresses and loads calculated for the 30 inch FDTS upper weld am bounding. Physically, the upper and lower welds of the TS sleeve are a duplicate of the FDTS upper weld. The FDTS sleeves at 30 inches are
)
subjected to much higher loads than the shorter 9.0 inches or less TS sleeve, and themfore, -
the following analysis for the FDTS sleeve bounds the TS sleeves.
l~
i i
i I
l l
i i
l I
l i
l l
1 Report No. CEN-629-NP, Revision 03-NP Page 8-3 1
--=
l ABB Combustion Engineering Nuclear Operations
{
I TABLE 8-1
SUMMARY
OF RIREVE AND WFI.n ANALYRIE SIGNIFICANT RF_RULTS -
l 1
i
-[CAMY ALIDWARIE*
ANALYSIS RESIILTS.
IECA11DN i
N.!!@ l M/, <
s/$si)J -
' (maxj(Stress islai))
R ' (MM 1
t Primary Stmss S = 26.6 General S.I. = 15.2 Sleeve j
(Normal Design) j
- @@PM[ $ PATS Slesse Wald(0.000inchM)W '
- W9' * -
Pnmary local Stress 1.5 S = 40.0 Stress Intensity = 14.2 Across Sleeve 1
Stress Range 3 S = 80.0 Stress Intensity = 36.0 Outside Sleeve I
Fatigue U = 1.0 Usage Factor = 0.09 Outside Sleeve i
Main Steam Line 0.7 S, = 56.0..
Stress Intensity = 24.3 Sleeve l
Break (MSLB)
(Accident Cond.)
t 84
1 Stress Range 3 S = 80.0 Stress Intensity = 46.5 Top of Weld j
i Fatigue U = 1.0 Usage Factor = 0.33 Top of Weld i
.6(.7S,)= 33.6**
Shear Stress = 22.4 Weld l
Break (MSLB)
(Accident i
Cond.)
2 K g-f S' immerSeeve Wald(FlyrS) y $.
" ' [, 7 l
4 Primary Iax:al Stress 1.5 S = 40.0 Stress Intensity = 27.9 Inside Sleeve j
Stress Range 3 S = 80.0 Stress Intensity = 40.9 Inside Sleeve l
Fatigue U = 1.0 Usage Factor = 0.27 Inside Sleeve l
Prirnary Pipe Break 0.7 S, = 56.0**
Stress Intensity = 11.4 Sleeve (LOCA)
(Accident Cond.)
{
i i
-The allowables listed in Table 8-1 are in wd.i with the ASME Code (Refs. 8.1 and 8.15).
-The mimmum tensile strength is listed in Reference 8.15 as 80.0 kai. This value will be used to evaluate the l
accident condssons and the allowable sleeve wall degradation in Secuan 8.3.
[
i f
i i
Report No. CEN-629-NP, Revision 03-NP Page 8-4 i
l i
1 1
ABB Combustion Engineering Nuclear Operations i
FORMULAS FOR GENERAL MEMBRANE STRESSES SUMMARIZED IN TABLE 8-1 (Note: All SI equations below are a derivation of the formula in Par. NB-3324.1 of Ref. 8.1.)
- 1. GENERAL PRIMARY MEMBRANE STRERS (DESIGN TUBESHFFT A PRFRSURE) j S.I.,,,,,,,,,
P R P
+-
P = AP = 1600 psi t
2 i
S.I.,,,,e = (l.600) (.333) + 1.600
.037 2
J S.I.e = 15.2 ksi < Allowable of S,,, = 26.6 ksi i
- 2. MAIN STEAM IJNE BRFAK S.I.usw = APR + AP t
2 1
Where, AP = 2.560 ksi (Per NRC Generic letter 95-05: " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking", Page 3 of Attachment 1, as applied to the Westinghouse Plants) i S.I.scu = (2.56) (.333) + 2.56 i
i
.037 2
S.I.,m = 30.6 ksi < Allowable of0.7 S, = 56.0 ksi i
- 3. PRIMARY PIPE BREAK (LOCA)
S.J.un = APR + AP t
2 Where AP is the maximum secondary side hot standby pressure (-1.085 ksi, external in Reference 8.2), which is less than approximately 7.7 ksi required for instability failure to occur i
with this type of external pressure application. Thus, the equation for internal pressure is j
applicable for this AP external pressure value.
i S.I w c4 = (-1.085)(.370), (-1.085)
S. I.wc, = -11.4 ksi < Allowable of0.7S, = 56.0 ksi i
Report No. CEN-629-NP, Revision 03-NP Page 8-5
ABB Combustion Engineering Nuclear Operations TABLE 8-2
SUMMARY
OF LOWER JOINT (WFi.nED AND ROI I Fn) DESIGN. ANALYSIS AND TEST RE9ULTS
)
l
- n.,i&yu::
0 bh
(;Dlh... k. ?wl h.
WIkkhh]
n &.c g A. R,4Y%
Mb f
N?f$ ;-
j jg.,
1 l [ f f.[_Q:].2l
'y,,; y) 1 AXIAL LOAD 2
1 Normal Ooerafmg
[
] - max. for operating plants with
[
]
welded / welded design by analysis (pg. 8-17) no weld joint slippage i
[
] - max. for operating plants with
[
]
welded / rolled design by analysis (pg. 8-17) no rolled joint slippage l
[
] - max. for " worst" case envelope
[
]
with welded / welded design by analysis (pg. 8-no weld joint slippage 18)
[
] - max. for " worst" case envelope
[
]
3 with welded / welded design by analysis (pg. 8-no weld joint slippage j
18)
Accident MSIR
[
] by analysis (pg. 8-7)
[
]
l no rolled joint slippage i
FATIGUE Zero Slippage - By Design no weld or rolledjoint slippage or leakage 3
Report No. CEN-629-NP, Revision 03-NP Page 8-6 a
I ABB Combustion Engineering Nuclear Operations l
i
,5 8.2 ~
LOADINGS CONSIDERED
!/
l
_In this section a number of potential failure modes are examined to determine the relative l
safety margins for selected events.
Failure loads are calculated based on minimum dimensions and compered with mechanical testing results from Section 7.0. Both calculated
{
and measured loads are compared with the maximum postulated loads.
-8.2.1 Unner hve Weld PnHmu I nmi t
For the purposes of this calculation, a weld height of 0.080" is assumed. This value reflects the actual minimum weld height that has.been observed by ABB-CE during qualification testing and process improvement testing. This value exceeds the required minimum weld height of 0.020" developed in Section 8.3.3, but better reflects the actual weld condition.
Assuming the parent tube is totally severed, the minimum load required to shear the upper l
tube weld is calculated..The forte required to pull the expanded sleeve through the
{
l-unexpanded tube is conservatively neglected.
1 1
I I
l i
l l
l l
In the event of a main steam line break (MSLB), the pressure differential would be 2560 psi i
per NRC Generic letter 95-05: " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking", Page 3 of, as applied to the Westinghouse Plants. Postulating a main steam line break
_ MSLB) accident, the maximum pullout load would be:
(
i Report No. CEN-629-NP, Revision 03-NP Page 8-7 L
y j ' la -
y.-
.n, - -.,
ABB Combustion Engineering Nuclear Operations 8.2.2 Imwer Sleeve Rolled or Weld Rection Puchnut I nad Assummg the parent tube is totally severed, the minimum load required to rupture the lower rolled section is calculated. The minimum measured test value for the pushout load is
~
[
] for the weld section. See Section 7 for details.
Postulating a loss of primary coolant accident (LOCA) during hot standby condition (0%
Power), the maximum available load would be:
Note that the LOCA pipe bmak accident is not controlling for this joint. See Section 8.4.6.
i 8.2.3 Weld Fatigue l
Since the factors of safety are quite high for loadings due to primary stress, the failure mechanism of greatest interest is the fatigue failure mode considering the variable axial l
!aading of the sleeve during normal operating transients.
In Section 8.6, fatigue evaluations of the upper weld and of the lower weld, which join the '
sleeve to the tube will be made. It is first necessary to determine the effects that tube lock-up within the tubesheet and tube supports have on the axial loads in the sleeve during normal operation. This subject is addressed in Section 8.4.-
l i
8.3 EVALUATION FOR ALLOWABLE SLEEVE WALL DEGRADATION USING REGULATORY GUIDE 1.121 j
NRC Regulatory Guide 1.121 (Reference 8.3) requires that a minimum acceptable tube (or j
sleeve) wall thickness be established to provide a basis for leaving a degraded tube in l
service. For partial thru-wall attack from any source, the requirements fall into two
)
categories, (a) normal operation safety margins, and (b) considerations related to postulated pipe rupture accidents.
8.3.1 Normal Oneration Kafetu Margins It is the general intent'of these requirements to maintain the same factors of safety in evaluating degraded tubes as those which were contained in the orig' al construction code, m
' ASME Boiler and Pressure Vessel Code,Section III (Reference 8.1).
l For Inconel Alloy 600 and 690 tube or sleeve material the controlling safety margin is:
Report No. CEN-629-NP, Revision 03-NP Page 8-8
ABB Combustion Engineering Nuclear Operations
" Tubes with partial thm-wall cracks, wastage, or combinations of these should have a factor of safety against failure by bursting under normal operating conditions of not less than 3 at any tube location".
From Reference 8.2, the normal operating conditions for the " worst" case envelopment of steam generators are:
Primary Pressure Pg = 2250 psia Secondary Pressure P,c = 653 psia Differential Pressure DP = Pg - P,c = 1597 psi Average Pressure P,,, = 0.5 (Pg + P,c) = 1452 psia Assuming the parent tube is totally severed, the sleeve is required to carry the pressure loading. The following terms are used in this evaluation.
Rs = sleeve nominalinside radius Sy,, = mimmum required yield strength (per U.S. NRC Reg. Guide 1.121)
Sym, = mmimum yield strength of sleeve (Sy = 35.2 ksi min. at 650 F)
Based on the information provided in Table 8-1, the sleeve material has a mmimum tensile strength of 80.0 ksi at 650 F. The required thickness is shown below using a dcrivation of the formula in Paragraph NB-3324.1 of Reference 8.1 with 3 times AP as mentioned in j
)
ReguRegulatory Guide 1.121 and S, in place of S.
m Therefore, the [
] allowable degradation is controlling for the normal operating conditions.
i Report No. CEN-629-NP, Revision 03-NP Page 8-9
ABB Combustion Engineering Nuclear Operations 8.3.2 Pmentated Pine Runmre Accidents i
NRC Regulatory Guide 1.121 requires the following:
"The margin of safety against tube failure under postulated accidents, such as a IDCA, f
steam line break,- or feedwater line break concurrent with the safe shutdown canhquake
-(SSE), should be consistent with the margin of safety determined by the stress limits specified in NB-3225 of Section III of the ASME Boiler and Pressure Vessel Code".
l The above referenced ASME code paragraph deals with " faulted conditions", where for an elastic analysis ofInconel 690 sleeves, a general membrane stress of 0.7 S = 0.7(80.0) =
o 5fia ksi is allowed. In conjunction with the NRC Regulatory Guide 1.121, the following 1'
accidents are postulated:
I (a) For a downcomer feedrmg steam generator, a feedwater line break (FWLB) accident l
would have very little effect on tube spans just above the tubesheet.
l (b) A LOCA accident causes large tube bending stresses in the upper tube bundle but produces only negligible compressive stresses in the mgion of interest.
(c) The only significant loading from a main steam line break (MSLB) accident on a sleeve is the differential pressure which is' considered on page 8-5. If the sleeved tube is locked into the first tube support, there is a potential for a small amount of
. additional axial stress. However, this axial stress would have a small value and was therefore neglected due to the following considerations:
a.
Based on ABB-CE's past experience with the calculation of flow loads due to a j
steam line pipe break, it is not certain in which direction the flow would load the lowest tube support. Cases have been observed, in which the flow path of least resistance was downward through the first tube support and then upward through the downcomer annulus.
b.
Since any load on the tube support would be shared with other locked tubes, the load per tube would be small (less than 10 pounds), based on an EPRI study performed by ABB-CE. (See Reference 8.8).
The required thickness for a main steam line break (MSLB) accident is shown below using
_ the derivation of the formula in Paragraph NB-3324.1 with.7 S, in place of S.
m l
1 i
Therefore, the [
] allowable degradation in Section 8.3.1 is controlling for both the l
main steam line break (MSLB) accident and normal operatirg conditions.
Repon No. CEN-629-NP, Revision 03-NP Page 8-10 l
ABB Combustion Engineering Nuclear Operations 8.3.3 Minimum Weld Heiaht 12enuirement There are significant margins available in the calculations using the upper weld length of
.080". The following calculations determine the nunimum required length of the upper sleeve weld for Design and Faulted conditions assuming the tube is severed. The required height is shown to be.02" for Design Conditions and.015" for Faulted conditions. These required sizes satisfy the ASME Code requirements ( NB-4357)fe similar type welds.
Design Condition:
i
)
'\\
l
_ Faulted Condition - Main Steam Line Break:
]
i i
i I
i Report No. CEN-629-NP, Revision 03-NP Page 8-11 L
ABB Combustion Engineering Nuclear Operations 8.4 EFFECTS OF TUBE LOCK-UP ON SLEEVE LOADING Objective: Conservatively determine the maximum axial loads on the sleeve (tension and i
compression) during normal operation.
l t
General Assumptions: (See Figures 8-2 through 8-5 for details).
1.
The model is a system of axial members with properties and boundaries in Tables 8-3A and 8-3B.
2.
Point B is fixed (tube in tubesheet and secondary face of tubesheet).
l 3.
Locations C and D are rigid (sleeve to tube weld and sleeve to tube rolled joint).
4.
All adjacent tubes are unplugged, unsleeved and locked between the tubesheet (point B) and the first tube support above the upper sleeve weld joint.
5.
Member 3 (tube inside the tubesheet) is locked into the tubesheet at both ends and is, 3
therefore, forced to move as the tubesheet moves.
l 6.
The axial load results from using very conservative secondary side temperatures based l
on a circulation ratio of 2. The actual circulation ratios are near 4.
i k
i l
t I
t i
i
)
l 8.4.1 weved Tuhe in Oneratino Reanm Generninr. Free at Tube Sunnort
),
i Assumptions: Point A is free. The sleeve is 30 inches long and inserted near the tubesheet primary face (Figure 8-1). The axial load is a function of the spring constants for the sleeve and tube lengths, the location of the sleeve's lower (rolled or welded) joint in the tubesheet, and the forced displacement between the sleeve and tube.
l f
Report No. CEN-629-NP, Revision 03 NP Page 8-12
ABB Combustion Engineering Nuclear Operations i
)
The thermal growth of the three axial members is represented by 6, the stiffness by K:
Li ar (Trn - 70) 61
=
62 = La a2 (Ta - 70) 63 Ls as (Tra - 70)
=
l Bru,wi 62 + Ss - 6:
=
1 1
K2 since member 3 is rigid K2s
=
j i
-+
+0 K2 Ks K2
)
i l
The sleeve loads, F, are in Table 8-4A for the transient conditions shown in the same table.
i l-l Thus:
6runa Ki
, butfor equilibrium F2 I'
- Fi
=
Kr
)
1 7herefore: 6,m = (F /K;) + (F;/K) i 2
Sem = F *(K; + K)/(K,K) 2 2
2 Transposing:
Fi = Staa
- Ki + K2 8.4.2 heved Tube in " Worst" Cnw Envelonment. Free at Tube hnnort Assumptions: Point A is free. The sleeve is 30 inches long and inserted near the tubesheet primary face (Figure 8-1). The axial load is a function of the spring constants for the sleeve and tube lengths, the location of the sleeve's lower (rolled or welded) joint in the tubesheet, and the fomed displacement between the sleeve and tube. The thermal growth of the three axial members is represented by 6, the stiffness by K. The developed equations are the same as those in Section 8.4.1. The sleeve loads, F, are in Table 8-4B, using the transient i
conditions in this same table.
Report No. CEN-629-NP, Revision 03-NP Page 8-13
i
' ABB Combustion Engineering Nuclear Oper:tions l
t p
3 8.4.3 Waved Tube in Oneratine me=m Generator. I mk-un at First Tube hnnort f
I
' Assumptions: Point A is locked-in to the first tube support above the upper sleeve weld f
joint and is therefore, forced to move by surrounding tubes. The sleeve is 30 inches long (Figure 8-1). (Hypothetical only, not currently locked).
{
i The composite member, CD or 6, is the assembly modeled in Figure 8-5.
i b
Ke = K1 + K2 l
Se = (61 + Ar) net elongation (Results in Table 8 - 4A)l
)
So = Ls as (To - 70)
Bs = Ls as (To - 70)
I The term Sw, can be calculated from data in Tables 8-4A and 8-5A. The following Sw.s = E + E, butfor equilibriunt Fs = Fe Ks Ke Therefore: Buna = E + &, Ke Fe
- Ks Fe, p Ms + Ke)
Ks Ke
. KsNe KsNe KsNe Transposing:
Fe = &n.a
- Ks + Ke relationship can be developed from the model depicted in Figure 8-5:
Then,' the deflection of memher 6 is A = F /lQ, and, referring to Figure 8-4, the deflection 6
of member 1, the sleeve, is A = F /K. The sleeve load is therefore:
i i i F = K x (As + A )' where A is from Table 8-4A.
The axial forces, i
i 6
i F, are in Table 3-5A for the transient conditions given in this same table.
i u -
8.4.4 heved Tuhe in " Worst" Onee Enveinnment I mk-un at First Tube hannrt Assumptions: Point A is locked-in to the first tube support above the upper sleeve weld joint and is therefore, forced to move by surrounding tubes. The sleeve is 30 inches long (Figure 8-1).'(Hypothetical only, not currently locked). The developed equations are the same as those in Section 8.4.3. The term, Sw, can be calculated from data in Tables 8-4B and 8-5B. The axial forces, F, are in Table 8-5B for the transient conditions in this i
same table.
Report No. CEN-629-NP, Revision 03-NP Page 8-14
TABLE 8-3A 30 INCH SI.FEVE - AXIAL MEMBER PHYSICAL PROrEx11ES FOR OPERATING S'fEAM GENERATOR NMYTbhlif)@ N.
-i$[I 05b 2I^
76 ! M 'CUBP[
k*
h k$ :I.^
j j,g}
'm
- p.pgg 4 D {5 pj
}. ggfg
';; u Y? W 3( )
.II' Vgf M(%d;
@e$nw%s.x w e..
jfQ;W.Q i;;glNigf ~ :jl,y
% ?
"i MM 4MW~
g at S. W,"_ + wx n ~?=
y--
in 1
a w m. i.64c m
x w
. _n n sms c.
.x Reference Temperatures:
Pnmary Giot) = 594T Secondary = 46TF Normal Tubes = (2 Tg + T )/3 = 551.TF m
NOTE:
' Nominal Dimensions for sleeve from Reference 8.9.
~
2 a and E for Inconel 690 from Ref. 8.14, Part D, Tables TM4, TE4 (same or more conservative than Ref. 8.11).
m
' Nominal Dimensions for tubes from Reference 8.13.
- u and E for Inconel 600 from Reference 8.14, Part D, Tables TM4, TE4 m
5 a for Carbon Moly Steel from Reference 8.14, Part D, Table TE-1.
m 8-15
ABB Combustion Engineering Nuclear Operations TABLE 8-3B 30 INCII SLFFVE - AXIAL l\\fEMBER PHYSICAL PROPERTIES FOR " WORST" CASE ENVELOPMENT n m, n.
- i. n u e,,
=.,
v 3 LOUTsfDd.: dNSIDE,5 M W,_;, wN.SBCTidN =w< a WMm,k blEAN;COEFt u
., @ 5 RAD d ;MW 3h 3?dRRidh $M
'& i_g1 '
Y r
FM M/W1 r
al-vi(. M T3.g, W g?fsyU$ i ygm - Epfi ~ 7?E k S y' fg_ g?
i s 1rtilo;+
M m$ f 4 ~) shfinE_@~T,idfin,*FEl0?~?
w ma w.
W en
-- q m %.r a
s
., m. ote :m i
w e
o,.
N(in)S3 Vf(i.n)B, ~U $..
[1: sW 4 d_ f M;(inb74
~.
i%
n, ~,a na Primary (Ilot) = 594T Secondary = 46TF (506T') Normal Tubes = (2 Tg + T )/3 = 551.TF(564.TF")
Reference Temperatures:
m NOTE:
Nominal Dimensions for sleeve from Reference 8.9.
2 ot,,, and E for Inconel 690 from Ref. 8.14, Part D Tables TM-4, TE-4 (same or more conservative than Ref. 3.11).
3 Nominal Dimensions for tubes from Reference 8.13.
- ot,,, and E for inconel 600 from Reference 8.14, Part D, Tables TM-4 TE-4.
5 ot,,, for Carbon Moly Steel from Reference 8.14, Part D, Table TE-1.
6 Values for the central bundle region, used in Appendix 8B calculation.
Report No. CEN-629-NP, Revision 03-NP Page 8-16
ABB Combustion Engineering Nuclear Operations TABLE 8-4A 30 INCH SI.FFVE -
AXIAL LOADS IN SI.FFVE WITH TUBE NOT LOCKFn INTO TUBE SUPPORT FOR OPERATING STEAM GENERATOR Q.~
- W' :-Ll -
.,,,y f-jg =
.jg 3g y
.fp f4
^2 R
=
~N fw$?
im m% a 3, x us ;,,,
- e y
-n >,
u w:g.ga;
- qqyLy c
, m
-1 i;g&syg; w-n u
gg g
- www - u;e gy
.w
. _:u n,o
. wna vn,g
- 1:c l
l NOTE:
'Due to small variation, E and ot, value for normal operation,100% power, are used.
i Report No. CEN-629-NP, Revision 03-NP Page 8-17
ABB Combustion Engineering Nuclear Operations TABLE 8-4B 30 INCH SI.FFVE -
AXIAL LOADS IN SI.FFVE WITH TUBE NOT LOCKED INTO TUBE SUPPORT FOR " WORST" CAW ENVEI OPMENT NWfk;f :
? 45Y YEY? $f):--fpkh$
,.l $Nh W? $h%Q'l l^
NQ p - T I:
- [s%
_ NOTE:
'Due to small variation, E and a value for normal operation,100% power, are used.
m 2Values for central bundle region, used in Appendix 8B calculation.
I Report No. CEN-629-NP, Revision 03-NP Page 8-18
ABB Corabustion Engineering Nuclect Operations TABLE 8-5A 30 INCH SIFEVE -
AXIAL LOADS IN SIFEVE WITH TUBE LOCKED INTO TUBE SUPPORT FOR OPERATING STEAM GENTRATOR Surrounding ~
- Upper Composim '-
Composite -
. Resuhant -
' Ibbes Tube '
- Member' Member -
Sleeve Imi
- Ts T.
T, = (2T,+T,)/3
. Deflection Deflechon b
- -Isad Load Deflecnon Deflecuen F = D '. K -
i i " i M
d.
..F.
Ds _= Fv1C, ds ~
D ' = D, + D.
4 c
i
' On)
(Ibs)
(F) -
' (F)
(F) '
On)
- On)
= Gn).
_Obs)
~ Gn)
.c NOTE:
'Due to small variation, E and a, value for normal operation,100% power are used.
n Report No. CEN-629-NP, Revision 03-NP Page 8-19
ABB Combustion Engineering Nuclear Operations TABLE 8-5B 30 INCH SIREVE -
AXIAL LOADS IN SIRFVE WITH TUBE LOCKED INTO TUBE SUPPORT FOR " WORST" CASE ENVELOPMFNT M-Surrounding Upper 4
_ Composite Composite
' Resultant
'Ibbes '
- Tube Member
' Member '
Sleeve Load -
T..
Te = (2T,+T,)/3.
. Deflection
. Deflection 6
- . / Imad ~-
. load Deflection,
. Deflection.
F = 4*K.i i
_ CONDITION.
6 (F)
- (F)
(*F) ^
(in)
(In)
(1#
' (Ibs) '
. (In)
A*. = 4 + A -
6 -.
4
.j.
, _ F.
'A = FgIQ -
3 (In) '
(Ibs)
NOTE:
'Due to small variation, E and ot, value for normal operation,100% power are used.
Report No. CEN-629-NP, Revision 03-NP Page 8-20
ABB Combustion Engineering Nuclear Operations 8.4.5 Effect of Tube Prestress Prior to Sleeving If a tube becomes locked into the lowest tube support, it can develop an axial preload.
The lock-in is more likely to occur during normal operation. It might be thought that a lock-in tube is always under significant preload during a shutdown.
This is not necessarily the case. The support behaves, for the most part, like a beam on an elastic foundation. It offers little resistance to tube axial growth except near a vertical tie rod or a connection poin; to the flow shmud. Reference 8.8 studied this behavior in some detail.
A preloaded tube is not likely to have a significant effect on the cyclic performance of a sleeve welded to it. While a tube loaded to the point of yield or buckling would cause some redistribution of the load in an installed sleeve, the sleeve loading range would stabilize on a few cycles and then shakedown to elastic action.
1 Therefore, the prestressed state of a locked-in tube to be sleeved is not of significant concern, so long as it does not hamper the sleeve installation process.
n 8.4.6 Lower Sleeve Rolled or Weld Section Pushout Due to Rectrairwl Thermal Ermncion In Section 8.2.2, the potential for lower sleeve rolled or weld section push-out due to LOCA is considered. However, the maximum compressive load in the sleeve due to thermal expansion is calculated to be [
] (maximum) for operating plants and [
] (maximum) for " worst" case envelopment as compared with [
] due to LOCA.
Therefore, comparing the thermal expansion load with the test value of [
] for the rolled joint and [
] for the weld joint (Section 7.0), will yield the controlling factor of safety.
l
)
Report No. CEN-629-NP, Revision 03-NP Page 8-21
h ABB Combustion Engineering Nuclear Operations i
8.5 SLEEVED TUBE VIBRATION CONSIDERATIONS l
The vibration behavior is reviewed since the installation of a sleeve in a tube could affect
)
l the dynamic response characteristics of the tube.
f.
8.5.1 Effects ofIncreased Stiffness i
Stiffness and mass have opposing influences on tube vibration. While increased stiffness i
tends to raise the tube natural frequency, increased mass tends to lower it. ABB-CE's l
vibrational testing (Reference,8.6) demonstrated among other things, that a solid rod of the same O.D. as a tube will vibrate at nearly the same frequency. However, the
{
displacements for the stiffer rod will be significantly less.
-l I
l In additiori, if any contact is made between the tube and sleeve along their length, the l
increased damping will absorb more energy. The damping would have a significant j
l effect on amplitude of vibration. In light of this damping effect and the other above j
mentioned effects resulting from a sleeve inside a tube, the vibration performance of the tube / sleeve assembly is superior over the original tube.
l l
l 8.5.2 ' Effect of Severed Tube i'
A severed tube (100% degradation defect) held together by a tube sleeve is a difficult geometry to analyze from a dynamic response viewpoint, and it is therefore more reliable and practical to obtain required information from testirig.
l One test program included evaluating the Palisades tube sleeve design (Reference 8.16).
{
Dynamic response tests were performed to ascertain the effects of the tube sleeves when -
l exposed to the vibration characteristics of a typical Palisades steam generator tube array.
j l
L The Palisades sleeves were different from the tube sleeves considered in this report in
.that they were hydraulically expanded in the tubes rather than welded, Also they were totally above the tubesheet and were not designed to be leaktight but rather " leak resistant".
The tube sleeves in this evaluation are designed to be leaktight.
The
. increased " tightness" of this design makes the sleeve less susceptible to vibration damage.
1 i
The following discussion is a summary.of the Palisades tube sleeve vibration tests and results.
1 I:
L r
Report No. CEN429-NP, Revision 03-NP-Page 8-22 l
i
m.
t t
- L l
ABB Combustion Engineering Nuclear Operations l
l l
l f
b l
I
?
I I
a i
h f
i
?
i i
i i
t i
Y
?
i l
i i
l i
i l
I.
i l
l 1
i I
i i
)-
Report No. CEN-629-NP, Revision 03-NP Page 8-23 i
\\
t ABB Combustion Engineering Nuclear Operations I
8.6 STRUCTURAL ANALYSIS FOR NORMAL OPERATION A static elastic analysis of the sleeved tube assembly was performed according to the requirements stipulated in NB-3220 Section III of the ASME Code Section. This section describes the methods used to analyze the upper tube weld and lower stub weld.
l 8.6.1 Fatione Evahi=* inn of Unner Tube /cleeve Weld j
e The Finite Element Method (FEM) was incorporated in this analysis, using the ANSYS Computer Code (Reference 8.5). Figure 8-6 depicts the FEM model of the upper tube -
The lower end of the tube was assumed to be locked near the secondary side surface of j
the tubesheet.
From Section 8.4, it was found that the sleeve develops higher compressive loadings if the tube is free to slide through the first support. Therefore, sliding at the tube-to-support interface was conservatively assumed. The FEM model consists of 2-D isoparametric elements with an axisymmetric option. The ANSYS input and output data are included in Attachment 1.
j The transient conditions listed in Reference 8.4 are shown in Table 8-6 and are grouped as follows for simplicity of analysis:
The 200 cycles between ambient (room temperature) and hot standby represent e
the heatups and cooldowns.
)
The 20,500 cycles between hot standby and full power are the sum of 18,300 j
loading and unloading conditions and 2200 step load events.
)
i The 600 cycles between full power and reactor trip are a combination of 400 trip, e
80 loss of flow, 80 loss of load and 40 loss of power cycles. "less of Flow",
which is assumed to represent the greatest variation from full power, is utilized to define the " Trip" condition.
The axial loads determined from the thermal interaction in Section 8.4 are applied to the bottom of the sleeve FEM model. The pressure stresses and stresses due to radial thermal expansion are conservatively excluded since they result in tensile stresses which relieve the compressive stresses resulting from the axial loads. The above described transients are combined in the worst case combinations in the fatigue evaluation.
12ak test and hydro test are isothermal and produce negligibly small sleeve loads. The upper weld edges are insensitive to pressure cycling, hence, test conditions are not
. considered in the upper weld fatigue evaluation.
Report No. CEN-629-NP, Revision 03-NP Page 8-24 a
ABB Combustion Engineering Nuclear Operations i
At the. weld tip region, a stress concentration factor of four (4) is applied to the total stresses from the computer code output for the purpose.of calculating peak stresses. The results of the analysis, which consist of the nodal stress tabulations at the critical sections and fatigue usage factors, are contained in Appendix 8A.
The minimum required axial length of weld of.02 inches was determined in Section 8.3.3.. A fatigue analysis was performed for this configuration. The finite element model j
used for the.08 inch weld design was modified by refining the element mesh as shown in Appendix 8A. The stress concentration factor of four (4) is applied to the linearized j
stresses from the computer output. The results of the analysis are also contained on Appendix 8A. All stresses and usage factors for both configurations are satisfactory i
when compared to allowables.
8.6.2 Evninntinn of Lower Sleeve Rolled Section The lower section of the sleeve will be rolled into the tube for the sleeve design with welded upper joint and rolled lower joint. Normal operating and transient conditions used in the cyclic loading tests on the steam generator tube sleeves are based on the tubesheet flexure (internal pressure differential) and differential thermal expansion of the tube and sleeve.
The transient conditions listed in Reference 8.4 are shown in Table 8-7. The logic for this grouping is as follows:
i The 400 cycles between ambient and hot standby represent the 200 heatup and cooldown conditions entnbined with 200 primary side leak tests.
The 20,500 cycles between hot standby and full power are the sum of 18,300 I
loading and unloading conditions and 2200 step load events.
The 600 cycles between full power and reactor trip are a combination of 400 trip, l
80 loss of flow, 80 loss ofload and 40 loss of power cycles.
The 880 cycles between reactor trip and secondary leak test are composed of 800 tube leakage test conditions and 80 secondary side leakage tests. A pressure of 840 psi was conservatively selected to represent the secondary leak test condition.
The tubesheet ligament stress in the load cycling tests is based upon the maximum allowable primary membrane stress intensity of 1.5 S or 40 ksi for the tubesheet material for the maximum design tubesheet differential pressure, (i.e. AP = P - P ). Of i
2 1600 psi. For a rolled joint at [
). This ligament stress value is used in the load cycling tests. This test value is conservative when compared to the tubesheet ligament stresses in Figure 8-9 which come from Reference 8-12.
Report No. CEN-629-NP, Revision 03-NP Page 8-25
ABB Combustion Engineering Nuclear Operations TABI E 8-6 UPPER SITEVE WFI D - TRANSIENTS CONSIDERFD
- fjy
/RESTRAINEDI 4.Tk;
.... ~
jTHERMAI4 TRANSIENTS END POINTS TCYCIMi
? EXPANSIONS
- ( F)i
?("F)/
i lLOADSdbsE 233 Ambient 0
70 70 (1) Heatup/Cooldown 0% S.S.
200
-790 547 547 (2) Loading / Unloading 0% S.S.
-790 547 547 (0% - 100%)
100% S.S.
20,500
-1420 594 467 (3) Reactor Trip 100% S.S.
-1420 594 467 Loss of Flow 600
-538 500 540 CONDITIONS:
(a)
Worst case: (a) Tube is not locked-in to first tube support (b) Tube is near periphery.
(b) Tube is Intact: Tube / Sleeve restrained thermal expansion.
(c)
Pressure stress is not significant in fatigue.
(d) Temperatures are as stated in Tables 8-4A and 8-4B.
(e)
Axial loads are from Tables 8-4A and 8-4B.
(f)
Sleeve is 30 inches long.
(g) Maximum Loads for both tube sleeve designs at " worst" case envelopment.
i I
i Repon No. CEN-629-NP, Revision 03-NP Page 8-26
TABLE 8-7 LOWER SLEEVE SECTION - TRANSIFNTS CONSIDEREll
^mR RESTRAINEDil!
n
' ~
ik N -
m TRANSIENTS END POINTS MVrtkl THERM /EATANf ~ d!Ed #
$p$jf MS8b isi ~
furkifLOADh a hsi)}
si
,;-.om w
^
i[(Ibs)p
~
. Am s
(1) Ambient Ambient 0
0 0
0 (2) Hot Standby Ambient 0
0 0
0% S.S.
400
-775 2250 1085 1165 (3) Full Power 0% S.S.
100% S.S.
20,500
-1237 2250 720 1530 (4) Reactor Trip 100% S.S.
2 2
Loss of Flow 600
-524 1650
%3 687 (5) Secondary Leak Ambient 2
2 Test Test Pressure 880 0 (negligible) 0 840
-840 CONDITIONS:
(a)
Worst Case: Tube is not locked-in to first tube support.
(b)
Tube is Intact: Tube / sleeve restrained thermal expansion.
(c)
Differential pressure causes tubesheet flexure.
(d)
Axial loads are from Tables 8-4A and 8-4B.
(e) 200 Heatup/Cooldowns plus 200 primary leak tests equal 4DD cycles.
(f)
Sleeve is 30 inches long.
NOTE:
' Loads for the central tube bundle at " worst" case envelopment, used in Appendix 8B calculatica.
2 Reference 8-4.
Report No. CEN-629-NP, Revision 03-NP Page 8-27
.~.. _
i ABB Combustion Engineering Nuclear Operations
)
8.6.3 Fatione Evnheion ofImwer Stub Weld The lower section of the sleeve will be welded into the tube for the sleeve design with l
welded upper and lower joints. Figure 8-7 shows the finite element model of the tube to l
l sleeve weld. The model includes the weld region and the surrounding tubesheet ligament j
isoparametric elements for the entire region through the use of the ANSYS Computer j
Code (Reference 8.7). The Finite Element Method (FEM) is incorporated in this analysis using the FEM results from Reference 8.4.
The results are interpolated for the axial loads in this report's tube sleeves of 1237,775, and 524 lb at the central tube bundle j
location where the maximum tubesheet flexure occurs.
l The sleeve end on the design with welded upper and lower joints is force driven in the l
tube which results in a contact area in compression. The model is sufficiently refmed to i
develop concentrated stresses in the weld region which very closely approximate the j
actual stress state. With the design operating transients, the stresses due to various j
combination of differential pressure and thermal conditions are investigated. The flexure l
stresses in the tubesheet impose a uniform expansion or contraction, as shown in Figure 8-7, depending on whether the flexure stresses are tensile or compressive.
From i
4 previous experience, ABB-CE established the flexure stresses for a typical Westinghouse l
tubesheet at various operating transients. The resulting expansion or comraction is i
applied on the FEM model. The axial load due to restrained thermal growth is applied to 3
the sleeve end. Pressure loading shown in Table 8-7 is conservatively not applied to the l
model due to its overall effect on total stress and fatigue.
The transient conditions listed in Reference 8.4 are shown in Table 8-7. The logic for the grouping is as follows:
{
I The 400 cycles between ambient and hot standby represent the 200 heatup and e
cooldown conditions combined with 200 prirnary side leak tests.
The 20,500 cycles between hot standby and full power are the sum of 18,300 j
loading and unloading conditions and 2200 step load events.
l The 600 cycles between full power and reactor trip are a combination of 400 trip, i
e 80 loss of flow, 80 loss of load and 40 loss of power cycles.
l i
The 880 cycles between reactor trip and secondary leak test are composed of 800 e
tube leakage test conditions and 80 secondary side leakage tests. A pressure of 840 psi (Reference 8.4) is selected to represent the secondary leak test condition.
l The above described transient combinations are significantly conservative. Leak test and i
hydro test are isothermal and produce negligibly small sleeve loads.
A stress concentration factor of four (4) is applied to the linearized primary-plus-secondary l
stresses for purposes of computing the fatigue usage factors.
l
. Report No. CEN-629-NP, Revision 03-NP Page 8-28 i
ABB Combustion Engineering Nuclear Operations The results of the analysis,.which consists of deformed shape plots, nodal and element stress tabulations at the critical sections, and fatigue usage factor computations, are in Appendix 8B. All of the stresses and usage factors are satisfactory when compared to the allowable stresses. The highest stresses and usage factors occur across the tube / sleeve just above the tube-to-sleeve weld, rather than across the weld throat. Figures 8B-1 through 8B-4 illustrate the exaggerated deformations which occur due to forced displacement caused by tubesheet flexure and axial compressive load in the sleeve. The deformed shape is comparable to the failure.
i l
l i
r i
l Report No. CEN-629-NP, Revision 03-NP Page 8-29 1
i ABB Combustion Engineering Nuclear Operations i
8.7 REFERENCES
FOR SECTION 8.0 i
l 8.1 ASME Boiler and Pressure Vessel Code,Section III for Nuclear Power Plant Components,1995 edition.
I 8.2 ABB-CE Ixtter Report No. CSE-%-115, " Tube Sleeve History Data for 7/8 inch Steam Generator Tubes", May 03,1996.
l 8.3 U.S. NRC Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes".
8.4 ABB-CE License Report CEN-331-P, Rev.1-P, " Zion I & 2 Steam Generator Tube Repair Using leak Tight Sleeves", June 03,1986.
8.5 "ANSYS" Finite Element Computer Code, Rev. 5.1, 1994 by Swanson Analysis l
Systems, Inc.
8.6 " Vibration in Nuclear Heat Exchangers Due to Liquid and Two-Phase Flow," By i
W.J. Heilker and R.Q. Vincent, Journal of Engineering for Power, Volume 103, Pages 358-366, April 1981.
i 8.7 "ANSYS" Engineering Analysis System, User's Manual, Rev. 4.1, 1986, by John l
- A. Swanson.
i 8.8 EPRI NP-1479, "Effect of Out-of-Plane Denting 1.oads on the Structural Integrity of l
Steam Generator Internals," Contractor: Combustion Engineering, August 1980.
8.9 : ABB-CE Drawing No. C-SGN-218-058-01, " Welded Sleeve for 7/8" Diameter l
Westingiv-Steam Generator", January 1986.
l f
8.10 ABB-CE Drawing No. C-SGN-218459-04, " Welded Sleeve Installation Westinghanse 7/8" Diameter Tubes", February 1989.
8.11 Inconel 690, Huntington Alloys, Inc., Huntington, W. Virginia.
l 8.12 " Primary / Secondary Boundary Components Steady State Stress Evaluation",
Prepared by Raymond Paul Wedler, Westinghouse Electric Corp., April 1%5 l
(REF-96-001).
)
l-
[-
8.13 Westinghause Steam Generator Standard Information Package, January 04, 1982 (REF-96-002).
j 8.14 ASME Boiler and Pressure Vessel Code,Section II, Materials,1995 edition.
l l
i i
I Report No. CEN 629-NP, Revision 03-NP Page 8-30 L
I
i l
ABB Combustion Engineering Nuclear Operations i
j 8.15 ASME Boiler and Pressure Vessel Code Case N-20-3, "SB-163 Nickel-Chromium-Iron Tubing (Alloys 600 and 690)... at Specified Minimum Yield Strength of 40.0 ksi...,Section III, Division 1, Class 1", November 30,1988.
I 8.16 ABB-CE Report No. TR-ESE-178, Rev.1, " Palisades Steam Generator Tube / Sleeve i
Vibration Tests", October 05,1977 (REF-%-003).
1 i
i i
1 i
3 I
i 1
1 Report No. CEN-629-NP, Revision 03-NP Page 8-31
ABB Combustion Engineering Nuclear Operations I
~m
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FIGURE 8-4 STIFFNESS MODEL OF SLEEVE AND LOWFJt TUBE
- Report No. CEN-629-NP, Revision 03-NP Page 8-35
i ABB Combustion Engineering Nuclear Operations i
A E
Surrounding Upper 2
Tubes (5)
Tube (4)
K4 C e?
B Composite Member (6)
K6 Tubesheet (3)
Dm FIGURE 8-5 STIFFNESS MODEL OF UPPER TUBE AND SURROUNDING TUBES Report No. CEN-629-NP, Revision 03-NP Page 8-36
ABB Combustion Engineering Nuclear Operations
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i FINITE FIFMENT MODEL OF UPPER 'I1JBE WFT D Report No. CEN-629-NP, Revision 03-NP Page 8-37 l
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ABB Combustion Engineering Nuclear Operations
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ABB Combustion Engineering Nuclear Operations TUBE BUNDLE DESIGN AP
= 1750 PSIG PRIMARY PRESSURE 2520 PSIG
=
SECONDARY PRESSURE
=
770 PSIG PRIMARY TEMPERATURE
=
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=
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Repon No. CEN-629-NP, Revision 03-NP Page 8-39
5 ABB Combustion Engineering Nuclear Operations 4
l i
i i
I t
i l
e APPENDIX 8A i
i I
FATIGUE EVALUATION OF UPPER TUBE / SLEEVE WELD
(
1 I
i I
i i
i i
i t
Report No. CEN-629-NP, Revision 03-NP Page SA-1 i
4 r
z.
l ABB Combustion Engineering Nuclear Operations t
INTRODUCTION
-The analysis parei in this appendix is discussed in detail in Section 8.6.1 of this Report. The 1
- results from the two (2) finite element models considered are presented in this Appendix. The j
model geometry is shown in Figure 8-6 of the report. The only difference in the two models is
_j the weld height and the number of elements. The 0.080" weld height model is based on the design geometry minimum dimension. The 0.020" model is based on the minimum required i
axial weld length for ogidixig and accident conditions. All stresses and usage factors for both j
configurations are satisfactory when compered to allowables, j
GENERAL DISCUSSION j
i The lower end of the tube was assumed to be locked near the secondary side surface of the i
tubesheet.' From Section 8.4, it was found that the sleeve develops higher compressive loadings l
l if the tube is free to slide through the first support. Therefore, sliding at the tube-to-support j
interface was conservatively assumed. The FEM model consists of 2-D isoparw-uie elements i
with an axisyomma-ie option. The ANSYS input and output data are included in Attachment 1.
l The transient conditions listed in Reference 8.4 are shown in Table 8-6 and are grouped as follows for simplicity of analysis.
I The 200 cycles between ambient (room temperature) arul hot standby represent the heatups j
l and cooldowns.
l The 20,500 cycles between hot standby and full power are the sum of 18,300 loading and l
unloading corvlitions and 2200 step load events.
The 600 cycles between full power and reactor trip are a combination of 400 trip, 80 loss of
)'
flow, 80 loss of load and 40 loss of power cycles. "less of Flow", which is assumed to represent the greatest variation from full power, is utilized to define the " Trip" condition.
6 The axial loads determined from the thermal interaction in Section 8.4 are applied to the bottom l
of the sleeve FEM model. The pressure stresses and stresses due to radial thermal expansion are conservatively excluded since they result in tensile stresses which relieve the compressive stresses resulting from the axial loads. The above described transients are combined in the worst j
case combinations in the fatigue evaluation.
1
)
i For the 0.080" weld model, a stress concentration factor of four (4) is applied to the total stresses from the computer code output for the purpose of. calculating peak stresses. The concentration factor is only applied to the axial and radial stresses since the shear stresses are j
relatively negligible. The concentration factor is applied at the sleeve outside surface located L
below the weld, the top and bottom of the weld, and to the inside surface of the tube at the i
location above the weld.
1 i
o r
1' Report No. CEN-629-NP, Revision 03-NP Page 8A-2 l
a
ABB Combustion Engineering Nuclear Operations l
t
}
For the 0.020" weld tip region the stress concentration factor of four (4) is applied to the j
linearized stresses from the computer output.
The factor is conservatively applied to the i
linearized membrane plus bending stresses for the axial, radial and shear stress components.
i i
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ABB Combustion Engineering Nuclear Operations l
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FIGURE SA-1 NODE AND STRESS CUT IDENTIFICATION Report No. CEN-629-NP, Revision 03-NP Page 8A-4
t 1
e ABB Combustion Enginee:ing Nuclear Operations l
TABLE 8A-1A i
i i
STRESS RESULTS 100% STEADY STATE
- 1
.I j
i SLEEVE, SECTION BELOW WELD (KSI) t Q:cmutsg B Q
r
,4
+79 U"Axiol:
07,
=
Oasehd
- 0 O
m
'3.
Hosp ( 4 l
E l
4 l
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t i
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WELD (KSI) j O
F i"JV A
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Gamed
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3*
Hoop?
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i TUBE, SECTION ABOVE WELD (KSI) 4&L(X:AMONg j
( *%gg (gi "lgh Report No. CEN-629-NP, Revision 03-NP Page 8A-5
ABB Combustion Engineering Nuclear Operations TABLE 8A-1B
\\
STRESS RESULTS 0% STEADY STATE l
1 SLEEVE, SECTION BELOW WELD (KSD J
a
,,a
)
b;:.:
- f Ushdial WN
~ IGA sik'
- $ 'G k J '
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WELD (KSO i
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,,g, cy:ggg.
n wn
.n,z
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- n w:
p,,ggg, n
,77e gg
, f yy p.
2 sy g
Report No. CEN-629-NP, Revision 03-NP Page 8A-6
i l
ABB Combustion Engineering Nuclear Operations e
ii-i TABLE 8A-1C J
1 4
i STRESS RESULTS REACTOR TRIP I
i a
i SLEEVE, SECTION BELOW WELD (KSI)
-ef a
- .s e s.
.va.
s.
g-,.
I, Gamm, [l 1h
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.r#'
yg,,[h.
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-i WELD (KSI) i i
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m.
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a.
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4 4
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[
f[
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[h:t pfGh.
~,16h.
i v
Report No. CEN-629-NP, Revision 03-NP Page 8A-7
--._.._________._.____-__.__m.~.___.___._
.l ABB Combustion Engineering Nuclear Operations
)
TABLE 8A-2 1
4 4
FATIGUE EVALUATION AT WORST LOCATION 4
SLEEVE, SECTION BELOW WELD - OUTSIDE SURFACE 3
STRESS INTENSITIES i
l
- PRINCIPAL STRESSES (KSI) iSI$fb f SI21 ~ 'IISI5 ?
~
Okamais, UIxist
,UHoopk (W
n, j
i.
i i
f 1
t i
s 8
OUISIDE SURFACE - SI, (SLEEVE, SECTION BELOW WELD)
'I h
Ma,iCONDTHONS)
_ SM
,~Sc w Syi_
N4;, :, N k ?"
6U-
)
1--
i 1
l l
1 i
4 k
4 4
\\
4 1
}
By inspection of Tables 8A-1A,1B and IC it can be seen that the usage factor for the other j
l locations will be less than.09.
l 1
1 4
4 I
i I
t i
l 1
4 Report No. CEN-629-NP, Revision 03-NP Page 8A-8 i
2
ABB Combustion Engineering Nuclear Operations t
i 4
lII
- +-
1 lf Eli siceve H
iit li i i i i i i i
l i
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ll l l lll l l l l l l l
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)
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I l l I H I iiil l II l l
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Weld Region 66 I i ! 1 I 11 i i i i i i i
l l
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I I
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i
~
1 FIGURE SA-2 NODE AND STRESS CUT IDENTIFICATION FOR 20 MIL WELD Report No. CEN-629-NP, Revision 03-NP Page 8A-9 1
4 j.
ABB Combustion Engineering Nuclear Operations TABLE 8A-3A t
t STRESS RESULTS 100% STEADY STATE (0.020" Weld) i 2
e i.
j SLEEVE, SECTION BELOW WELD (KSI) i iY;g'"' MW,}p;{,;.+i u.-
a<
...s.
,1 j :
yp a.-
, 4); gg^
,5.. -, -
.+.
~,,s
..%QLe j'i.l-t " ' Q,,,,.,U e s 9 4; '.i%
f;
~O#
- ')
,l'
!G
}$_% ;. # 1
- /"U
. M AnM iic ? ~
ileep~.;
hi u
+
r a
1i l
1 h
i..
3 I
1i 3
[
WELD (KSI) l Qu@mm% 3auy
- Jak fug$ %%
I TUBE, SECTION ABOVE WELD (KSI) 6xMiggyy vW ~
- Q ~fa.wsdg > % Ge$,e.
4.;, xw ', W@-..sn;G :4xs -w_nTener:n u
ns WMGashs RE agempWP s ec '
- ~'
Report No. CEN-629-NP, Revision 03-NP Page 8A-10
q ABB Combustion Engineering Nuclear Operations 7ABLE 8A-3B STRESS RESULTS QS STEADY STATE (0.020" Weld)
SLEEVE, SECTION BELOW WELD (KSI) 6jLOCATIONz
- G Q g
. Um.
m =. G >
4 Tm.
m 44, WELD (KSI)
.y J LOCA, TION ~;
- U M
^
4
~-
~;.;g
>G m-
< Gw;;
LT m TUBE, SECTION ABOVE WELD (KSI) tG m" M
.f. : G m,s;.
'a LOCA, TIONJ s
_M a:;. > 'G ' '; ;7 3m-:.
+
m F
l Report No. CEN-629-NP, Revision 03-NP Page 8A-11 g
y:
~. _ _ _. _ _ _ _ _ _ _ _ _. _ _...
i t
l ABB Combustion Engineering Nuclear Operations o
I TABLE 8A-3C l
t i
STRESS RESULTS REACTOR TRIP (0.020" Weld) i t
d b
i SLEEVE, SECTION BELOW WELD (KSI) l
.i i
- A LOCATIONM
!. C 7,m m.
ou Sheer :
l 5'-
4
> a w %.:
. gnamdi; -
V. <:' g Ax W e c,
igHoop,
t I
i t
i 1
1 a
I i
4 I
i l.
1-t WELD (KSI) l l
LOCAHONa-
- U z gRamal l
j
?r
,,: g H'oopf:
cfe 1 & sar.
AnW
+
,e j
.i I
J s
i
{
4 l
1-i i
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I i
i i
i TUBE, SECTION ABOVE WELD (KSI) l u
.. mygy c re,; Gam:m,eu_,
+0
- w.,
l
. b; O +o c.H spi-4y8 hear.-
l
-c.
.q
.n
,-z AuWV A
~
t i
i 4
1 1
4 a
t 1
i i
i i
i i.
E i
4 1
4 e
t j
Report No. CEN-629-NP, Revision 03-NP_
Page 8A-12 i
i 4
i l
1 i
ABB Combustion Engineering Nuclear Operations l
TABLE 8A-4A RANGE OF STRRE AT WORST LOCATIONS (0.020" Weld) i 1
SLEEVE, SECTION BELOW WELD - OUTSIDE SURFACE x y :p
, gg ggg, g, ggg w;.'cf'nhso w ~ ~,,
ig:: CONDITION,
o
> :Sll"
- S2'-
S3 A J,1-S3h
'S243(
- S142
i Max. SI Range = [
] < 3 S = 80 ksi I
i 1
?
WELD,1DP t
i;%rgn 4 v.., ',- "...
" NAL mRM %
6 MN '
2 h
- ,gPODBU) MON',
9 S1H
- S2l fS3Q ';S143)
S243 e:S142?'
l i
i Max. SI Range = [
] < 3 S, = 80 ksi i
i i
l i
t Report No. CEN-629-NP, Revision 03-NP Page 8A-13 i
{
i ABB Combustion Engineering Nuclear Operations -
1
)
TABLE SA-4B i
FATIGUE EVALUATION AT WORST LOCATIONS (0.020" Weld) i i
i i
SLEEVE, SECTION BELOW WELD - OUTSIDE SURFACE 1
I
.,..f8 f
PRINCIPAL STRESSES' PEAK STRESS INTENSITIES' Z
{
w A. c,im i
,, ' If f.
4"'#'
y.
..S1?w
-- S2 e I
. 3:CONDITKINk
.!S3 ;
- SI-S3F
< 'S2 83 g S1-S2 l, 4
/
1-i I
l I
I OUTSIDE SURFACE - SI (SLEEVE, SECTION BELOW WELD) j f
yxjCONDm0NSie-,;1SM S#,.
Se -
2M
[N,:,:
'" %U::
N u
i I
\\
NOTE: A stress concentratie.,n factor of 4 is applied to the axial, radial and shear stress.
1 j
i l
1 l
l i
i 1
I
\\
Report No. CEN-629-NP, Revision 03-NP Page 8A-14 l
l
.p..
-y 7
-__,----7
5 ABB Combustion Engineering Nuclear Operations TABLE 8A-4B (Cont'd) i FATIGUE EVALUATION AT WOltisT LOCATIONS (0.020" Weld)
?
wED, MP Arg JV.
PRINCIPAL STPRtWEX.
~
At
- u
, W PEAK STRESSINTENSITIES T l
71, x,gg, j.i
.Sl M
S3[
TSIM s D-3'14 'JSI-S22 4
J I
i i
1 i
o.
i i
^
i i
WEw, MP 1
tc[.jCONINHONS 9
- S.P
- ) Sg,
LSJK
. :NA;,
', h N4 iU:
't
~a s
4 j
1 o.
I i
i NOTE: A stress concentration factor of 4 is applied to the axial, radial and shear stress.
By inspection of Tables 8A-3A,3B and 3C it can be seen that the usage factors for the other locations will be less than.33.
i i
i i
t I
Report No. CEN-629-NP, Revision 03-NP Page 8A-15
]
i ABB Combustion Engineering Nuclear Operations i
4 i
j r.
h I
i i
1 APPENDIX 8B 3
FATIGUE EVALUATION OF LOWER STUB WELD l
i l
i i
Report No. CEN-629-NP, Revision 03-NP Page 8B-1
ABB Combustion Engineering Nuclear Operctions 1
DERIVATION OF EXPANSION / CONTRACTION ON LOWER TGRF RI FFVE WFI n JOINT DUE TO TURFRHFFT FI FXURE STRRRRES As discussed in Section 8.6.3, the tubesheet ' flexure stresses impos.e significant expansion / contraction on the tube to sleeve weld at the lowerjoint.
The average deflection will approximately be:
S, = (So + Sg)/2 = (R*(1 - v*)/2E*)(og + og) m R = 0.77 in, as shown in Figure 8-7.
E* and v* are the equivalent solid plate properties of the tubesheet.
og ami og are the equivalent solid plate stresses.
For this report, the sleeve location was conservatively at the center of the tubesheet since the flexure stresses are the highest at the center. The values of E*, v*, o, and o are obtained from o
n Reference 8.12. The Finite Element Method (FEM) incorporated in this appendix uses the FEM results from Reference 8.4 with results mterpolated for the axial loads from the central tube bundle at " worst" case envelopment. The rest of this appendix deals with the appropriate f' mite element plots and results.
l
.l l
Report No. CEN-629-NP, Revision 03-NP Page 8B-2
ABB Combustion Engincering Nuclear Operations AXIAL LOAD DUE TO RESTRAINED THERlW. EXPANSION 7
i
,,, i g
h a,
i
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il Ii j
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il l l alli f
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4
~,
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H
~~
w.r > /
C0t9RESSIVE
.+ -
-1 FLEXURE OF
. SLE. -
TE"..lLD ( HOT STANDBY )
TUBE 3HEIT i
E _._
FIGURE 8B-1 LOWER STUB WELD MODEL (HOT STANDBY)
Report No. CEN-629-NP, Revision 03-NP Page 8B-3
_ ___.. _..~.. _.___..__ _. _ __.._.._.
ABB Combustion Engineering Nuclear Operations I
AZIAL LOAD OUE M ltESTRAINED l
i THERMAL EXPAltSION I
i i
ai L
i w
I l
i l ti
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FLEXURE OF COMPRESSIVE i
s.
w a
b, g
TUBE 5HEET
.SLE(.
_yLD (FULL POIER & THERtPL LOAD)
FIGURE 8B-2 I
LOWER STUB WELD MODEL (FULL POWER & THERMAL LOAD)
Report No. CEN-629-NP, Revision 03-NP Page 8B-4
j-ABB Combustion Engineering Nuclear Operations AZIAL LOAD DUE i
TU RESTRAIMED THERMAL EXPANSION 1
'lr J
~
li i
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p i
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1
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l-d'll i
Il l
l ll il l I _
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l 4 $ 4 ! .' f.,
. / s-COMPRESSIVE t
'! 'd' FLE.%)RE OF l
, '. " W.
TUBE 5HEET u
S L _.+
JELD (REACTOR TRIP & THERt1AL LOAD) w
)
FIGURE SB-3 i
)
LOWER STUB WELD MODEL (REACTOR TRIP & THERMAL LOAD) i L
Repon No. CEN-629-NP, Revision 03-NP Page 8B-5
- ~
. ~. -
ABB Combustion Engineering Nuclear Operations 1
i i
l l
( \\\\
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' +:
- ?,: ;y Tasn.E asnmE W" f 0F TUBES'iEET 4
r E
/
SLEEVE--TUBE WELD- (SECONDARY LEAK TEST)
FIGURE 88-4
^
LOWER STUB WELD MODEL (SECONDARY LEAK TEST)
Report No. CEN-629-NP, Revision 03-NP Page 8B-6
~.
- - ~.-. -... -
ABB Combustion Engineering Nuclear Operations 1,
TABLE 8B-1 l
STRESS RESULTS i
SECTION 1 IN THE TUBE SLEEVE 1
i 1.5 S Stresses & S.I.(ksi)
.c
? c/;
' hl
- c.(("
- ogo,';
, o,-a,
, i 'ah,'.
Candhian j
i 1
i i
a Max. S.I. = [
] ksi < l.5 S. = 40.0 ksi i,
t
,1
}
SECTION 1 IN THE TUBE SLEEVE 3.0 S. Stresses & S.I.(ksi) enadinan R
- o,..
o, i'
' a, o,-o,a o,.o,,
c,-o..
4 i
i i
Max. Range of S.I. = [
] ksi < 3 S = 80.0 ksi I
l i
(
i i
P I
f Report No, CEN-629-NP, Revision 03-NP Page 8B-7 4
ABB Combustion Engineering Nuclear Operatic 1 1
i TABIR SB-2 FATIGUE EVALUATION SECT 10N 1 IN THE TUBE SLEEVE
~
(ksi) randman
~ S *,
. sg S-Nm N,,,., '
'U.
I l
I 1
l Total U = 0.270 < 1.0 J
- - with a stress concentration factor of four (4) 1 A
i i
i i
j i
i i
I
)
?
i i
.e I
l I
l i
l l
i i
Report No. CEN-629-NP, Revision 03-NP Page 88-8 1
li
4 ABB Combustion Engineeritig Nuclear Operations.
- ~
9.0 SLEEVE INSTALLATION VERIFICATION 9.1
SUMMARY
AND CONCLUSIONS l
The ABB CENO welded sleeve installation process and sequence has been tested to :
ensure the installation of a sleeve which conforms to the design criteria described in i'
Section 3.
During this testing, actual steam generator conditions, such as the influence of tubes locked at tube supports, have been considered in assessing the acceptablity of the various processes and the sequence in which they are performed.
1 l
Actual sleeve operating history, as well as the qualification test program described j
l within this report indicate that the ABB CENO steam generator tube sleeve is capable of performing as well as, if not longer than, the original tube in which it has been installed.
9.2 SLEEVE-TUBE INSTALLATION SEQUENCE i
9.2.1 Full Denth Tnhecheet Riceve with Rnlied Imwer Joint I
The FDTS Sleeve with the rolled lowerjoint is described in Section 4.3 and Figure 44.
Installation is accomplished' using the processes described in Section 4.5 in the following sequence-1 1)
Tube I.D. Cleaning i
I 2)
Sleeve Installation and Expansion 3)
Stmetural Weld Near the Sleeve Upper End 4)
Ultrasonic Examination of Sleeve Upper Weld
- i i
5)
Visual Eramination of Sleeve Upper Weld (Optional)
- 6)
Postweld Heat Treatment of Sleeve Upper Weld (Optional) l 3
7)
Sleeve lower End Torque Roll 8)
- Sequence may be performed interchangably 9.2.2 Tube Enanort sleeve The TS Sleeve is described in Section 4.3 and Figure 4-7. Installation is accomplished using the processes described in Section 4.5 in the following sequence:
/
- 1). Tube I.D. Cleaning
- 2)
Sleeve Installation and Expansion 3)
Stmetural Weld Near the Sleeve Upper End
.4)
Structural Weld Near the Sleeve lower End 5)
Ultrasonic Examination of Sleeve Upper and Lower Weld *
'6)
Visual Examination of Sleeve Upper and lower Weld (Optional)
- Report No. CEN-629-NP, Revision 03-NP -
Page 9-1
ABB Combustion Engineering Nuclear Operations 7)-
Postweld Heat Treatment of Sleeve Upper and Lower Welds (Optional) l 8)
- Sequence may be performed interchangably 9.3 WELD INTEGRITY Initiated in 1983, ABB Combustion Engineering has conducted a comprehensive development program to ensure weld joint integrity.
Tube I.D. brushing tests, sleeve-tube expansion tests and weld parameter evaluation tests were all completed as part of the process verification.
9.3.1 Cleaning ountincasinn In preparation for welding, any oxide layer must be removed from the tube I.D.
surface.
Extensive tests were performed to develop, qualify and improve the cleaning process and tool. The cleaning program that was completed resulted in the qualification of an air powered, expandable wire brush.
To satisfy cleaning requirements, the brush head must rotate at a minimum speed of 2000 rpm for a minimum of two passes over each weld area.
A test program was conducted to establish the reduction of tube wall thickness after cleaning with the abrasive brush. Four oxidized tube samples were cleaned using a qualified bmshing tool and process. It was determined that less than.0005 inches of metal was removed from each of the samples.
An additional test was conducted to determine whether the I.D. tube bmshing would introduce noise interference on the bobbin coil eddy current test. A clean section of tubing was baseline tested to determine I.D. noise levels. The tubing was subsequently air oxidized to produce an oxide layer on the tube. One half of the tube section was then brushed to remove the oxide coating and the sample was retested with the bobbin coil. The results showed that the oxide layer does in fact generate a l
noise component. However, after the tubing is bmshed, the noise level returns to that of the baseline data.
9.3.2 F=ncian Onali6ca'ian j
An extensive test program was performed to qualify the bladder expansion tool and process, which provides a tight sleeve-tube fit up in preparation for welding. This j
program considered tubing with thick, thin and nominal walls as well as tubing with different yield strengths.
l Report No. CEN-629-NP, Revision 03-NP Page 9-2
4 ABB Combustion Engineering Nuclear Operations -
Y j
9.3.3 Weld Onnlification i
A development program was conducted to ensure weld joint integrity. Sleeves were y
welded into steam generator tube mockups to confirm that the welded joints were leaktight. Weld parameters were established by performing a series of welds 'and sectioning the welds to ensure that sleeve design requirements were met. Additional l-welds ;were made where welding current, sleeve and tube wall thicknesses were varied. A test matrix was used to verify the sleeve installation with welding process h
parameter. tolerances.. The weld joints were made with varying sleeve and tube diametral expansions and weld tip locations.
The test program confirmed that j
allowable tolerances do not affect weld integrity.
i Sleeve welding is' qualified using an approved test procedure (Reference 9.6.7). The sleeving test procedure is in compliance with applicable sections of the ASME Code.
3 Sleeve weld operators are qualified using test records in accordance with applicable i
sections of the ASME Code.
f The test procedure specifies the requirements for performing the welds, the changes in essential variables which require requalification, the method for examining the test assemblies and the requirements for qualifying the weld operators. Sleeve welding is l'
' qualified by performing six consecutive welds which meet specified design
[
requirements. Weld operators are qualified by performing one successful weld.
i A development program examining the effects of secondary side conditions upon j
welded sleeves in steam generator tubes was conducted. Oxidized and non-oxidized l
steam generator tubes were sleeved using the same process parameters for each L
specimen type and the results showed no significant effect caused by secondary side conditions. The program also demonstrated that the cleaning procedure described in l
this report was effective in preparing the inside diameter of the tube for welding, e
The development program consisted of several tests as described below:
I In early 1984, ABB-CE size steam generator tubes (3/4 inch O.D. x.042 inch wall) l oxidized in a model boiler (5 months All Volatile Treatment (AVT) plus 30 parts per i
million chlorides,1.5 months: wet layup, and 2.3 months AVT) were abrasively brushed on the inside surface, then welded using Alloy 690 sleeves. This group l
. included two oxidized and two unoxidized specimens. Visual and metallographic i
1 examination indicated that no effect was caused by the secondary side condition.
f' A second group of ABB-CE size steam generator tubes from the same model boiler F
test was assembled in a tubesheet mockup to determine the effect of secondary side conditions and tube proximity. Fabricated oxidized tubes with I.D. cleaning were j
compared with virgin material. Visual and metallographic examination of the welds
- showed no significant differences and welding was not affected when tubes were i
i i
Report No. CEN-629-NP, Revision 03-NP Page 9-3
ABB Combustion Engineering Nuclear Operations welded in an array. An acceptable weld was also demonstrated after application of approximately 0.1 inches of moist magnetite on the tube O.D.
A tube section which was taken from a Ringhals Unit 2 steam generator was abrasively I.D. cleaned, and an upper weld joint was made. The joint was visually and metallographically examined. There was no difference between that joint and other joint specimens which were made from non-oxidized tubes.
9.3.4 Ultrasonic Testing Ouali6 cation Ultrasonic (U.T.) techniques are employed to confirm the presence of weld fusion into the tube. A test program was completed by ABB-CE to qualify the Ultrasonic Examination of sleeve-tube upper welds. Fourteen sleeve-tube weld specimens were prepared for this qualification program. Each weld was ultrasonically inspected and then hydrostatically tested to confirm U.T. results. Test results indicate complete correlation between ultrasonic and hydrostatic testing.
9.3.5 Postweld Heat Trent Ouali6 cation The tubing used in some steam generators has been shown to be very susceptible to the effects of Primary Water Stress Corrosion Cracking (PWSCC). As a result, these utilities must minimize the residual stress induced in the steam generator tubing associated with any repair process. If sleeving is selected as the repair method, the sleeve to tube weld joint as well as the weld heat affected zone and primary pressure boundary portion of the tube expansion requires annealing to minimize residual stresses. The Electric Power Research Institute (EPRI) has documented (Reference 4.7.4 of Section 4.7) evidence in support of the annealing process. It was determined thata{
] was required to maximize tube life.
ABB-CE has developed a computer controlled and powered heat treatment system utilizing a resistance heater. A test program was conducted to verify the acceptable performance of the heat treatment system in meeting the established EPRI guidelines.
Specific tests were performed to determine the axial loads induced in locked steam generator tubes caused by the annealing operation. Additional tests determined the variation in steam generator tube O.D. annealing temperature as a result of the following conditions: first time heat treatment cycle vs. subsequent heat treatment cycles for a given tube; tolerances on the elevation of the heater tool as delivered by the tool installation drive system; and the effect of adjacent tubes to absorb heat from the tube being annealed (black body test).
In order to minimize the amount of distortion and residual stress, the heater has been specifically designed to heat the shortest length of the tube possible while ensuring that Report No. CEN-629-NP, Revision 03-NP Page 9-4
ABB Combustion Engineering Nuclear Operations I
j-the weld, weld heat affected zone, and the tube expansion transition area outboard of
)
the weld joint is heated within the specified range.
1 i
During this development effon, the tests of greatest significance were those associated j
with determining the residual stresses in the tube both prior and subsequent to the postweld heat treatment. Generally two types of tests were performed to assess the effects of the heat treatment on the sleeve-tube joint. One. consisted of accelerated corrosion testing of the joint in an aggressive environment during which standard specimens would be used to benchmark time to failure with total stress. These tests, I
j described in Section 6.2.1, were intended to identify the highest localized stresses associated with tube expansion and weld shrinkage and verify the temperature / time a
parameters for the process. The second test method was to instrument long sections of i
i tubes in which sleeves were installed and heat treated and measure the strains, loads and deformations associated with the installation.
4 9.3.5.1' Instmmented Analysis oflocked Tubes l
)
Although the tubes in the accelerated corrosion tests had been locked at their supports i
when the sleeves were installed and heat treated, no lockup loads were utilized during the corrosion tests. In order to assess these loads, two tests were conducted. The first
)
involved investigating the effect on a typical tube in a Westinghouse Series 51 steam i
generator. A shonened tubesheet sleeve was installed in a 0.875 inch O.D.x 0.050 l
inch wall tube which was then insened in a load frame as shown in Figure 9-1. The test assembly represented a tube which had been locked at the first support plate. A full length (49 inch) sleeve-tube assembly was positioned in the load frame with the cross heads locked. Two LVDTs were positioned midway between the weld and the cross heads to measure the radial deflection of the tube. In addition an extensometer I
was used to measure the localized axial strain for a four (4) inch gage length centered on the weld.
i A plot of the temperature profile and the axial load measured is shown in Figure 9-2.
)
The results of this test are shown in Table 9-1. Although no measurements were taken, i
no abrupt changes in the tube diameter were observed along the length of the tube. It was concluded that the deformation experienced by the tube would not be detrunental
. either to the installation process, i.e.. in preventing the tool from being removed, or to the long term performance of the sleeve-tube joint as described in Section 5.
A similar test was performed on a two by four array of 0.750 inch O.D. x 0.042 inch wall tubes arranged in a square pitch and supported as shown in Figure 9-3.
This configuration replicates the first three hot leg supports of a typical Westinghouse D3 Series generator while conservatively simulating aspects of an ABB-CE unit.
In addition, this configuration is conservative when compared to a Westinghouse Series 44/51 steam generator. Four of the tubes were locked at their support location (though not at the Flow Distribution Baffle) by tack welding in four locations. The other four were free from the tubesheet to Support Plate No. 8. Two Tube Support (TS) sleeves and a tubesheet sleeve were installed in each tube as shown in the figure. The tubes Report No. CEN-629-NP, Revision 03-NP Page 9-5
)
.~.
..~ - - ~_. -.. -. _ - -
f 4
l ABB Combustion Engineering Nuclear Operations i
j were instmmented with strain gages to determine the strain in tie tute outer fibers.
l During the heat treatment of each sleeve the strain in the tube was rworded. A load cell was used to determine the total load in the upper most section of tube. In the case l
of this mockup, the heat treatment commenced at the upper most weld and preceded l
toward the tubesheet. Both sleeve welds (where applicable) were treated prior to any l
j
. strain gage measurements. A typical temperature / time plot is shown in Figure 9-4. The l
j results of the test are shown in Table 9-2. As is expected, the more times the tube segment experiences the heat treat cycle, the greater the resulting residual stress.
l l
Examination of the tube surfaces in the vicinity of the welds indicated [
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l 9.3.6 Summary I
In summary, ABB-CE has conducted a comprehensive development and verification j
j program to ensure weld integrity of its leaktight sleeves. Experience has shown that i
j oxide layers as visually confirmed to exist on the steam generator secondary side do not affect weld parameters and the abrasive cleaning method described in the report is i
effective in preparing the tube for welding.
9.4 ROr.IFn JOINT INTEGRITY j
i A development program was conducted to ensure the rolled joint of the sleeve was leaktight and capable of withstanding the design loads. The sleeves were rolled into j
I mock-ups consisting of steam generator tubes which had been rolled into blocks simulating the tubesheet. The sleeves were then tested to confirm the rolled joint was leaktight both before and after cyclic load testing. Tests of the rolled joint were also conducted where process parameters such as torque, tube diameter and roll location relative to the nickel and metal oxide bands were varied. A test matrix was used to verify the sleeve installation with sleeve rolling process parameter tolerances. The test program confirmed that the rolled joint integrity is acceptable within the allowable rolling process tolerances.
9.5 COMMERCIAL SLEEVE INSTALLATION ABB-CE's commercial sleeving experience is shown in Table 2-1. The success rate for all installed welded sleeves is greater than 98%. Since 1985, no sleeve which has been ra**d based on U.T. and V.T. has been removed from service due to service related degradation. ABB-CE's sleeving operational experience is shown in Table 9-3.
Y Report No. CEN429-NP, Revision 03-NP Page 94
.'ABB Combustion Engineering Nuclear Operations q
l i
1 i
9.6 REFERENCES
FOR SECTION 9.0 9.6.1
" Test Report on Steam Generator Tube Cleaning for Installation of Welded Sleeves,"
l TR-MCM-126, December 1985.
l l
9.6.2
. "An Investigation of the Installation of Welded Sleeves in R.E. Ginna Tubing," TR-
. MSD-128, February 1986.
i 9.6.3
" Sleeving Centrifugal Wire Brush Development and Life Test Report," TR-ESE-705, i
May 1986.
l i
,9.6.4
- S.G. TSP /RTZ Sleeving-Tube I.D. Cleaning for 3/4 Inch O.D. X.042/.043' Wall l
Tubes," TR-ESE-860 December 1990.
l l
9.6.5
. " Steam Generator Sleeving - 3/4 inch Program, Bladder Expansion Pressure," TR-1 i
ESE-755, April 1988.
l
.9.6.6
" Steam Generator Sleeving - 3/4 inch Program, Qualification of RTZ and TSP Sleeve Expansion Tools and Bladder Life Test," TR-ESE-869, December 1990.
9.6.7
" Welded Steam Generator Tube Sleeve Semi-Automatic Gas Tungsten Arc Detailed Welding Procedure Qualification," Test Procedure 00000-MCM-050, January 1990.
l l
-9.6.8
" Ultrasonic Examination of 3/4 inch O.D. S.G. Tube to Sleeve Upper Welds," TR-
)
400-001, August 1988.
l 9.6.9
" Qualification of the Postweld Heat Treatment Tool for Westinghouse "D". Series
-l Steam Generators," 00000-ESE-830, March 1991.
9.6.10 '" Qualification of the Roll Transition Zone (RTZ) Sleeve Rolled Joint," 00000-ESE-J 826, May 1991.
9.6.11.
J
" Test Report To Support The Use Of A Rolled Lower Joint In Tubesheet Sleeves For l
Westinghouse Series "44" And "51" Steam Generators," CEN-631-P, October 1996.
j l
E 5
J i
j' Report No. CEN429-NP, Revision 03-NP Page 9-7 0
i ABB Combustion Engineering Nucleor Operations TABLE 9-1 O.875 O.D. SLEEVED TUBE PWHT DATA l
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l TABLE 9-2 l
0.7$0" O.D. SLEEVED TUBE PWHT DATA i
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Report No. CEN-629-NP, Revision 03-NP Page 9-8
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ABB Combustion Engineering Nuclear Operations i
r Load Cell su o
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22 3/4" i
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45 7B*
J Extensometer i
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Cyclander Load Frame FIGURE 9-1 0.875" O. D. LOCKED TUBE TEST TEST MOCKUP 1
Report No. CEN-629-NP, Revision 03-NP Page 9-10 j
ABB Combustion Engineering Nuclear Operations t
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FIGURE 9-2
' O.875" O.D. LOCKED TUBE TEST TEMPERATURE AND AXIAL LOAD PROFILE Report No. CEN-629-NP, Revision 03-NP Page 9-11 l
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ABB Combustion Engineering Nuclear Operations i
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4
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LOAD CELL
/
TSP 8
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TSP 5 f
STRAIN GAUGES A1 & A2 l
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STRAIN GAUGES t
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TUBESHEET' l
FIGURE 9-3 0.750" LOCKED TUBE TEST - TEST MOCKUP 1
Report No. CEN-629-NP, Revision 03-NP Page 9-12
ABB Combustion Engineering Nuclear Operations Key Tube Temperature ----
Controller Temperature a
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FIGURE 9-4 l
0.750" O.D. TYPICAL TEMPERATURE PROFILES Report No. CEN-629-NP, Revision 03-NP Page 9-13
ABB Combustion Engineering Nuclear Operations 10.0 EFFECT OF SLEEVING ON OPERATION Multiple plant specific analyses have been performed to determine the effects of installation of varying lengths and combinations of FDTS and tube support sleeves.
I Sleeve lengths and various combinations of installed sleeves were used to evaluate the effect of sleeving on the hydraulic characteristics and heat transfer capability of steam generators. Using the head and flow characteristics of the pumps, in conjunction with the primary system hydraulic resistances, system flow rates have been calculated as a l
function of the number of sleeved tubes and the types of sleeves installed. Similarly, i
curves are generated from calculations that show the percent reduction in system
}
flowrate as a function of newly plugged tubes (per steam generator). These curves are i
derived from plant specific information based on the following steam generator
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conditions :
l 1
Number Of Open Tubes Per Steam Generator i
e Number Of Tubes Sleeved i
Primary System Flowrate e
Primary Coolant Temperature i
' This information has been used to generate tables, such as Table 10-1, that provide hydraulic equivalency of plugs and installed sleeves, or the sleeve / plug ratio. Table 10-1 is provided as an approximation only and is based on assumed operating parameters j
and sleeve types for a Westinghouse Series 51 plant. It must be assumed that some variations in the sleeve / plug ratio will occur from plant to plant based on operating parameters and steam generator conditions.
l The overall resistance to heat transfer between the primary and secondary side of the steam generator consists of primary side film resistance, the resistance to heat transfer through the tube wall, and the secondary side film resistance. Since the primary side i
film resistance is only a fraction of the total resistance and the change in flow rate is so small, the effect of this flow rate change on heat transfer is negligible.
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When the sleeve is installed in the steam generator tube there is an annulus between the sleeve and tube except in the sleeve-tube weld regions. Hence, there is effectively little primary to secondary heat transfer in the region where the sleeve is installed. The loss i
in heat transfer area associated with sleeving is small when compared to the overall length of the tube.
In summary, installation of sleeves does not substantially affect the primary system flow l
rate or the heat transfer capability of the steam generators.
h Report No. CEN-629-NP, Revision 03-NP Page 10-1 L
t i
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ABB Combustion Engineering Nuclear Operations TABI F 10-1 j
f IYPICAL SI FFVE TO PLUG EQUIVAT ENCY RATIO l
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CASE CONFIGURATION RATIO (Sleeve / Plug)*
1 FDTS (1)
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FDTS (1) + TS (1)
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3 FDTS (1) + TS (2)
[ ]
- This ratio should be considered approximate due to plant to plant variations.
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Report No. CEN-629-NP, Revision 03-NP Page 10-2 1