ML20085K931
| ML20085K931 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 06/21/1995 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20085K922 | List: |
| References | |
| ESR-95-00765, ESR-95-00765-R02, ESR-95-765, ESR-95-765-R2, NUDOCS 9506280182 | |
| Download: ML20085K931 (68) | |
Text
{{#Wiki_filter:,. .- [ li Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 2 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 1 of 27 1 l ESR 95-00765 EVALUATION - REVISION 2 g UNIT 1 CORE SHROUD EVALUATION BASED ON B110R1 INSPECTION RESULTS l LIST OF EFFECTIVE PAGES
- BPAGEj "On REVISION _
1 2 2-14 1 15 2 16-20 1 21 2 22-27 1 Attachment A 0 Attachment B 1 Attachment C 1 Attachment D 1 Attachment E@ 2
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l: . l Carolina Power & Light Company ESR # 95-00765 , Brunswick Nuclear Plant Revision 1 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 2 of 27 .i r TABLE OF CONTENTS Page # List of Effective Pages ......................................1 Ta ble o f Co n ten ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 EXECUTIVE SUM M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 PURPOSE............................................ 3 BACKGROUND ........................................ 3 C O N CLU SI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 EVA L U ATI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 D E S I G N I N P UTS . . . . . . . . . . . . . : . . . . . . . . . . . . . . . . . . . . . . . . . 4 SHROUD DESIGN ......................................4 SHROUD FABRICATION AND INSTALLATION . . . . . . . . . . . . . . . . . . . 8
, CAU S AL FACTO RS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 INSPECTION RES U LTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 RFO B109R1 ...................................... 9 RFO B 1 10R 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 A N A LY SI S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
SUMMARY
.......................................... 20 REFE RE N C ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 DOCUMENT UPDATES .............; ...................... 21 l ES R A CTI O N ITEM S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 l FIGURES (1) Reactor Vessel Cross-Section Showing Reactor Internals . . . 22 (2) Reactor Shroud Three-Dimensional View . . . . . . . . . . . . . . 23 _
, (3) Roll-Out View of inside Shroud Surface . . . . . . . . . . . . . . . 24 (4) Roll-Out View of Outside Shroud Surface . . . . . . . . . . . . . . 25 (5) Brunswick Shroud Plan View . . . . . . . ..............26 (6) Shroud Cross-Section Showing Welds . . . . . . . . . . . . . . . . 27 WELD H6b STRUCTURAL ANALYSIS . . . . . . . . . . . . . . . . . . . . . Attachment A S AFETY REVIEW PACKAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . Attachment B
SUMMARY
OF LOADS AND STRESSES AT WELDS . . . . . . . . . . . Attachment C GRAPHICAL
SUMMARY
OF H1 AND HS UT RESULTS . . . . . . . . . Attachment D l DISCUSSION OF INDICATIONS BELOW H1 WELD . . . . . . . . . . . . Attachment E l
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Carolina Power & Light Company ESR # 95-00765 l Brunswick Nuclear Plant Revision 1 i
;A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 3 of 27 Q EXECUTIVE
SUMMARY
PURPOSE i This ESR Evaluation is required as part of CP&L's commitment to USNRC Information ! Notice 93-79, " Core Shroud Cracking at Beltline Region Welds in Boiling Water l Reactors." As such, this evaluation accomplishes the following: ! i) Documents the in-Vessel Visual Inspections (IVVI) performed on the Core ! Shroud during Refueling Outage (RFO) B110R1. ,
- 2) Evaluates the current IVVI data relative to previous inspection results and analyses.
1
- 3) Provides justification to use the Core Shroud for a minimum of another two (2) I afel ra e n the present condition ing e next o fue cycl s (Cy l 10 and Cycle 11) without any operational changes or restrictions.] l :
BACKGROUND :._ i i in October,1990, RICSIL No. 054 reported cracking near the circumferential seam weld at the core beltline area cf the shroud in a GE BWR/4 located outside the United States. Based on recommendations contained in this RICSIL, the BNP Unit 1 shroud . 1 was inspected in July,1993, and a near 360* circumferential crack was confirmed on the inside diameter of the Top Guide Support Ring, at the weld to the shroud mid-section. EER 93-0536 (Reference 3) was issued in 1993 to assess Unit 1 shroud structural integrity. CONCLUSIONS The BNP-1 Refueling Outage (RFO) B110R1 Core Shroud inspections are complete , and evaluated in this ESR. This ESR concludes that structural integrity of the core shroud will be maintained, with full FSAR safety margins, for at least the next two fuel cycles (currently scheduled to end in April,1998), based on analysis of the
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f i l l Carolina Power & Light Company ESR # 95-00765 ' Brunswick Nuclear Plant Revision 1 ATTACHMENT TO ENGINEERING SERVICE REQUEST Page 4 of 27 e 4 inspection results. Future inspection plans will consider not only these inspection results, but will also consider continuing developments in the industry, to ensure utilization of the best information and technology to address the issue. Crack growth experienced during Cycle 9 was substantially iess than postulated by previous analysis. The inspection results from B110R1 show that the existing condition is essentially unchanged from the condition identified during the B109R1 I outage. Furthermore, the postulated crack lengths at the end of Cycles 10 and 11 are fully bounded by previous analyses and will not reduce the structural design l margins below allowable values. Therefore, the condition of the core shroud does not impose any restrictions to BNP-1 operation during the next two cycles. ; The BNP-1 Core Shroud is " acceptable as is" for Operating Cycles 10 and 11. EVALUATION DESIGN INPUTS
- 1. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel
' Code, Section XI,1980 Edition through Winter 1981 Addenda
- 2. Technical Specifications for Brunswick Steam Electric Plant Unit 1
- 3. UT and IVVI Inspection data from BMP-1 Refueling Outages 8 (B109R1) and l _
9 (B110R1) l SHBOUD DERIGN The reactor pressure vessel (RPV) was designed in accordance with applicable l portions of Section lli of the ASME Boiler and Pressure Vessel Code (B&PV Code), I l 1965 Edition through Summer,1967 Addenda (Reference 2). Although the shroud I i itself is not 3 Code component, the B&PV Code was used as the design basis for l determining limits for stress intensities. l The core shroud is a cylindrical assembly inside the reactor vessel, which provides
l , , . . Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 1 A TTA CHMENT TO ENGINEERING SERVICE REQUEST Page 5 of 27 \ 1 i a partition to properly distribute the flow of coolant delivered to the vessel. The safety design basis of the shroud is to: ' a) Provide a floodable volume in which the core can be adequately cooled in the ; event of a breach in the nuclear system process barrier external to the reactor vessel. i I b) Limit deflections and deformations of the reactor vesselinternals to assure that the control rods and the core standby cooling systems can perform their safety functions during abnormal operational transients and accidents. c) Assure that the safety design bases (1) and (2) above are satisfied so that the '
] safe shutdown of the plant and removal of decay heat are not impaired.
The core shroud is composed of three regions: an upper shroud which is bounded by the shroud head and the top fuel guide; a central region which surrounds the fuel; and a lower region which surrounds the lower plenum and is welded to the I re: Ior vessel shroud support ring. The three regions are of different diameters: the l top region is approximately 15'-9" diameter; the central region is approximately 14'-9"; and the lower region is tapered from44'-9" to 14'-3" (see Figures 1 and 2, and Reference 1). Roll out maps of the core shroud depicting the locations of l horizontal welds H1-H9, vertical welds V1-V11, and plates P1-P11 are shown in l ! Figures 3 and 4. The weld and plate designations were assigned for inspection l l purposes. A plan view at the top of the shroud, depicting the 36 sets of shroud l _ ] head bolt lugs is shown in Figure 5. I I The upper shroud consists of the separator support ring, the upper shroud cylindrical l shell, and the top guide support ring. The separator support ring is constructed from l l 6 ring segments having a cross section of approximately 6" X 6", cut from rolled l and annealed plate, welded together, then machined to final dimensions. The l separator support ring material is Type 304 stainless steel from a single heat, with I a carbon content of 0.078 wt%. Thirty-six pairs of shroud bolt hold down lugs are welded to this ring. This assembly is joined to the upper shroud shell at weld H1, which consists of a Double-J prep weld with a fillet on the inside. The shell is o formed from (2) 1 %" thick semicircular plates, welded together using a Double-U
f I . . Carolina Power & Light Company ESR # 95-00765 : Brunswick Nuclear Plant Revision 1 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 6 of 27 l prep. The carbon contents range from 0.049 - 0.060 wt%. The Top Guide Support Ring, with a cross section of 7%" X 3", is constructed and welded (H2) to the upper shroud shellin a manner similar to the separator support ring. The ring material has a carbon content range of 0.063 - 0.064 wt%. The 6 ring segments were fabricated from two heats of material. These welds are oriented such that the axial residual stresses pull across the short transverse orientation (end grain) of the ring material.
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The central region of the shroud consists of the mid-shroud barrel, the core support ring, and adjoining welds. The barrel is formed in the same manner as the upper shroud shell, but consists of three cylindrical sections joined together at welds H4 and H5. Carbon contents range from 0.048 - 0.064 wt%. The mid-shroud barrel is welded to the upper shroud assembly at H3, which consists of a Single-J prep weld l from the inside, with a back gouge and a fillet reinforcement on the outside. It is welded to the core support ring at H6a, which is a Double-J prep weld with a fillet reinforcement on the inside. The core support ring is similar to the separator and top guide support rings, and has a carbon content range of 0.063 - 0.067 wt%. The 6 ring segments were fabricated from 2 heats of material. l The lower region of the shroud consists otahe lower shell course, shroud support i ring, jet pump diffuser ring, and associated welds. The lower shell course is formed from (3) 1 %" thick plates welded together using Double-U prep welds to form a conical section. Carbon contents range from 0.053 - 0.058 wt%. It is joined to the core support ring at weld H6b, which is similar to H6a. The shroud support ring, _ which transfers the shroud weight and other loads to the reactor vessel, is 2" thick Inconel Alloy 600. The lower shell course is joined to the shroud support ring using a bimetallic Single Bevel prep weld with a backing ring on the outside (H7). The jet pump diffuser ring is also made of Inconel Alloy 600, and is joined to the shroud support ring at H8, and to the reactor vessel wall at H9. H8 and H9 are Double-J prep welds with fillet reinforcements. I Refer to Table 1 and Figure 6 for shroud weld and materials details.
l . - Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 1 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 7 of 27 l
'iTABLEll $ SHROUD, WELD DETAILS 1f '
COMPONENT DESCRIPTION 1 PIECE NUMBERS ; fMATERIALE ^ N lCOMMENTS$MIM '
< an'd WELD NUMBERt herr' WELD PREP? and CARBON WT% Y C~* ijMQ@gis$
UPPER SHROUD
! SEPARATOR SUPPORT Piece 5' 304 SS SSR assembled from six plate RING (SSR) 0.078 wt% segments, welded with 308 SS '
Double-U welds WELD H1 Double-J: , 308 SS ID welded first, OD back-Fillet on ID chipped, then welded . UPPER SHROUD Piece l' 304 SS Assembled from 2 rolled plates. SHELL COURSE 0.049 - 0.060 wt% > welded together by 308 SS e Double-U welds V1 and V2 WELD H2 Double-J; 308 SS 10 welded first, OD back-: Fillet on ID chipped, then welded -- TOP GUIDE Piece 6' 304 SS TGSR assembled from six plate SUPPORT RING 0.063 - 0.064 wt% segments, welded with 308 SS (TGSR) Double- U welds WELD H3 Single-J on ID; 308 SS ID welded first, OD back-3 Fillet on OD chipped, then welded J MID-SHROUD BARREL MID-SHROUD Piece 2* 304 SS Assembled from 2 rolled plates, TOP SHELL COURSE (Upper) 0.056 wt% welded together by 308 SS Double-U welds V3 and V4 WELD H4 Double-J 308 SS One of last two welds made to 3 assemble shroud. MID-SHROUD Piece 3* 304 SS Assembled from 2 rolled plates, { MIDDLE SHELL COURSE 0.050 - 0.064 wt% welded together by 308 SS Double-U welds V5 and V6 WELD H5 Double-J 308 SS One of last two welds made to assemble shroud. MID-SHROUD Piece 2* 304 SS Assembled from 2 rolled plates, LOWER SHELL COURSE (lower) 0.048 - 0.058 wt% welded together by 308 SS Double-U welds V7 and V8 WELD H6a Double-J: 308 SS Fillet on ID
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- Sun Shipbuilding & Dry Dock Fabrication piece reference numbers.
I . Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 1 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 8 of 27 l , ax , eTABLE!17 SHROUD, WELD: DETAILS 2 um , dyj);gfi$g COMPONENT DESCRIPTION ! PIECE NUMBERS? EMATERIAW . WiCOMMENTSSMRg'
^ end WELD NUMBER) fand WELO PREP? and CARBON WT% , ps, - ~ Di$$$7 CORE SUPPORT RING Piece 7' 304 SS CSR assembled from six plate (CSR) 0.063 0.067 wt% segments, welded with 308 SS Double-U welds , WELD H6b Double Bevel; 308 SS t
Fillet on ID LOWER SHROUD LOWER SHROUD Piece 4* 304 SS Assembled from 3 rolled plates, TAPERED SHELL COURSE V.053 - 0.058 m % welded together by Double-U
, welds V9, V10, VII WELD H7 Single Bevel on Alloy 82 root Gas Tungsten Arc Weided: ; ID; Fillet Welded (GTAW) Root: Shielded Metal - , Backing Ring Alloy 182 filler Arc Walded (SMAW) Fili [
on OD SHROUD SUPPORT RING N/A Alloy 600 Plate thickness is 2.0". ; (SSR) l WELD H8 Double-J Alloy 82 root GTAW root. with Fillets SMAW fill. Alloy 182 filler JET PUMP N/A Alloy 600 Plate thickness is 2.S*. ! _ DIFFUSER RING l WELD H9 Double-J Alloy 82 root GTAW root.- (attaches Jet Pump with Fillets SM AW fill.- Diffuser Ring to Reactor Alloy 182 filler Vessel) ,'
- Sun Shipbuilding & Dry Dock Fabrication piece reference numbers.
SHBOUD FABRICATION _ AND INSTALLATION The core shroud was designed by General Electric and fabricated by Sun Shipbuilding
& Dry Dock Company from January 1970 to November 1971. The core shroud was installed in February 1974, with fit-up and welding provided by Brown & Root. CP&L has performed a detailed review of the fabrication and installation records (Reference 1). No significant fabrication or installation details were discovered that would 1
indicate any material conditions unique from standard practice at the time of 4
, fabrication and installation.
l t {, s . Carolina Power & Light Company ESR # 95-00765 - Brunswick Nuclear Plant Revision 1
~A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 9 of 27 \
l 4 The weld material was 308 SS for circumferential welds H1, H2, H3, H4, H5, H6a, . l and H6b and vertical welds V1 - V11 and the Submerged Arc Welding (SAW) and/or i ! Shielded Metal Arc Welding (SMAW) processes were used. Welds H7, H8, and H9 l
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were made using inconel 82 and inconel 182 filler materials and the Gas Tungsten l ! Arc Welding (GTAW) and SMAW methods. l ; 4 i _[ CAUSAL FACTORR
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The factors that affect IGSCC and their relation to the core shroud are detailed in EER ; 93-0536, Revision 1 (Reference 3). This evaluation considered water chemistry, l } abroud materials and fabrication techniques, and critical hours of operation. Cracking histories of other components were also considered.
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t INSPECTION RERULTS i Inspections of various components within the reactor are routinely performed each refueling outage in accordance with the requirements of ASME Code Section XI, j vendor recommendations, and the plant in-Service inspection (ISI) Program. i RFO B109R1 ;._ The inspection plan for RFO B109R1 was based on the experience and observations from other BWRs and the current understanding of the causal factors contributing to l the cracking. Accordingly, the initial inspection plan provided extensive inspections I for the regions where the most cracking had been observed (above the core plate _ ; I (H1-H5) where neutron fluence and the oxidizing environment are most prominent), ! and sampled the regions where little or no cracking was present (below the core plate (H6-H9) and vertical welds). Evaluation and screening was performed in accordance with GENE-523-123-0993, Revision 2, " Evaluation and Screening Criteria for the Brunswick 1 Shroud Indications," dated 11/93 (Reference 5). Visual examinations involved cleaning the weld areas to remove surface film which might hinder detection of very tight indications. The distance to the shroud surface for visual examinations was established to discern a 1 mil wire (0.001 inch diameter) in order to ensure detection of tight cracks.
,, - - , . - - - - . . ~ .
I . - Carolina Power & Light Company ESR # 95-00765 i Brunswick Nuclear Plant Revision 1 ' A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 10 of 27 l l Ta'ble 2 provides a summary account of the inspection findings. Reference 8 l contains the specific inservice inspection results. LT BLh!22!
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(RESUt.TS5 T ma,s w a , a% a d i - H1 VISUAL INSPECTIONS: 100% of OD and ID surfaces were inspected. The cracks are long but not continuous. Primary orientation is circumferential, located on the OD, mainly below the bolting lugs in the Separator Support Ring. t Approximately 268* of the circumference (74%) has cracking. No consistent cracking pattern exists, except for some branching associated with attachment welds. Cracks have been found only on the outer surface. UT INSPECTIONS: Measurements made at 8 locations. Depths range from < 0.3* to 0.7 *. H2 VISUAL INSPECTIONS: 100% of OD was inspected. ID is not accessible. The cracks in the ; Top Guide Support Ring are long but not continuous. Approximately l 224* is cracked on the OD (62%). The cracks above the weld are ; small. LTT INSPECTIONS: Measurements made at 4 locations. Depths range from <0.3* to 0.7 5" . l I H3 VISUAL INSPECTIONS: 100% of OD and ID surfaces were examined. The crack is 360* l around the inside of the Top Guide Support Ring, except for gaps at i some of the Top Guide Support Ring's radial welds and some of the I eccentric aligner plate welds. The crack opening is more pronounced than for other cracks. The crack is approximately 1/16" from the toe l of the weld. - - l UT INSPECTIONS: Measurements made at 11 locations. Depths range from 0.95* to ) 1.71* ' H4 VISUAL INSPECTIONS: 100% of the ID and 45% of the OD were examined. Jet pumps prevented access to the remaining OD areas. Cracks are axially oriented (vertical). Cracks are located on both the OD and ID. Most of the cracks on the OD are located below the weld, while cracks on the ID are above the weld Cracking is concentrated at horizontal and vertical weld intersections. The longest crack extends less than 5' from the toe of the weld. Plates P3 and P4 exhibited a uniform distribution of axial cracking on the inside surface.
' I UT INSPECTIONS: None performed.
j (
I . l Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 1 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 11 of 27 I i: TABLE:2 - UNIT l'RFO'B109R1 DETAILED INSPECTION RESOLTSR ' e-
- WELD RESULTSi T
- ,,s H5 VISUAL INSPECTIONS: 100% of the ID and 45% of the OD were examined. 5 of the cracks on the OD are circumferential, extending from 0.5" to 3" in length.
The remaining cracks are axial, extending less than 9" from the toe of
} the weld. Long circumferential cracks (the longest is approximately 42.5") and short axial cracks (less than 4.5") appear on the inside surface.
UT INSPECTIONS: Baseline depth measurements taken in 2 locations to benchmark future inspections. Depth ranged from < 0.3" to 0.6* HGa VISUAL INSPECTIONS: 45% of the OD was examined, ID is not readily accessible. Only a few axial indications were found on the OD. The indications extend less
, and than 3" from the toe of the weld.
H6b UT INSPECTIONS: None performed. H7 VISUAL INSPECTIONS: 17% of the OD was examined, ID is not readily accessible. No indications identified. UT INSPECTIONS: None performed.
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H8 VISUAL INSPECTIONS: 17% of the OD was examined from the top side of the weld. ID and l bottom are not readily accessible. No indications identified. l UT INSPECTIONS: None performed. H9 VISUAL INSPECTIONS: 17% of the circumference was examined from the top side of the l
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weld. Bottom is not readily accessible. No indications identified. l UT INSPECTIONS: None performed, j V1 - V11 VISUAL INSPECTIONS: 100% of either the inside or outside surfaces of each vertical weld was examined. Additionally, the accessible portions of the other side were inspected. No indications were found on any vertical welds. I UT INSPECTIONS: None performed. PLATES VISUAL INSPECTIONS: All accessible areas of the ID and OD were inspected. One indication was found on the inside of Plate 6, at mid-plate, between welds H4 and H5. The mid-plate indication is oriented approximately 20' from horizontal, and is about 6.0" long. No other plate indications were found. UT INSPECTIONS: None performed. l
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Carolina Power & Light Company ESR # 95-00765 l Brunswick Nuclear Plant Revision 1 l A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 12 of 27 l l 2 i i
- lTABLEi2KUNIT 1 RFO:B109R1LDETAILED INSPECTION
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ATTACH. VISUAL INSPECTIONS: Additional inspections performed include: MENTS Top Guide Eccentric Pin and Brackets (4 locations)
' and -
Shroud Head Bolts (36 bolts) Shroud Head Bolt Lugs (36 pairs on the Head and COMPON- Shroud) ENTS - Shroud Guide Rod Bracket (O' and 180* azimuths)
. Shroud Head Guide Pin Bracket (0* and 180* azimuth)
Manway Access Covers (0* and 180* azimuths) Separator Support Ring Segment Welds (6 welds) 1 Top Guide Wedges (24 locations) l Top Guide Bolting (80 bolts around the periphery) Top Guide Hold-Down Latches (4 locations) , Top Guide Support Ring Segment Welds (6 welds) INDICATIONS 1 Eccentric Aligner Pin (180* location) ) 5 Shroud Head Bolt Lugs (6 indications totalin weld material) 1 indication bottom of the Top Guide (running from a bolt hole to a dowel pin hole at the 180* location).
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UT INSPECTIONS: UT inspedions were performed on the Manway Access Covers and no indications were noted. 9 Shroud Head Botts were l inspected and 7 found to be cracked. eaum I an-
4 1 - Carolina Power & Light Cornpany ESR # 95-00765 Brunswick Nuclear Plant Revision 1 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 13 of 27
,' RFO B110R1 The inspection plan was revised to exclude H2 and H3 since a qualified structural repair was performed during RFO B109R1. The plan was also modified to take advantage of weld specific analyses that had been performed - RAM-94-092/ SIR 029, " Addendum to the Brunswick Unit 1 Screening Criteria" dated 4/6/94 (Reference 6); and RAM-94-099/ SIR-94-031, " Minimum Required Unflawed Core Shroud Material at Brunswick, Units 1 and 2", dated 4/11/94 (Reference 7). Analysis indicated that the allowable length for an axial flaw exceeded the width of any of the plate material, so inspection of vertical welds was eliminated. l The inspection scope was structured to meet the intent of NRC Generic Letter 94-03, and focuses on three objectives:
- 1) Re-examination of selected areas to determine crack growth,
- 2) Examination of some of the installed clamps (spanning H2 and H3) to verify no inservice degradation, and
- 3) Utilization of specifically developed tooling to examine accessible portions of certain welds that could not be fully examined during RFO B109R1. l A sumrnary of the inspection plan is presented in Table 3, below.
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? TAB L,Et3% UNIT;11 RF.O: B 110R 1,#x. - CORE ~. H.ROUDllNSPECTIO.a 3:nN PLAN.iSUM
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%WELDb [~. Aj , j; ~ . s% INSPECTION METHOD AND SCOPE ;, X ',JIE.., Lpf N H1 VISUAL INSPECTIONS: None UT INSPECTIONS: Inspect four of eicht areas previously UT'd in RFO B109RI.
Inspections to be performed from the OD. H2 No inspections scheduled. Clamps installed during RFO B109R1. H3 No inspections scheduled. Clamps installed during RFO B109R1. 9
i-Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 1 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 14 of 27 yin
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?T BLEi3 U UNIT f RFO'BdOR1 CORE SHROOD ' INSPECTION PL a~ $UM$lIk.mm$$$
es n [ WELD j [lNSPECTION METH'OD ANDl$CNPEE; ' fj
, )f H4 Inspection not scheduled. No circumferentialindications were identified during RFO B109R1 inspections, so sufficient structural margins exist. Any anticipated growth of axialindications would not impact structural margins. Results of H5 inspections will be considered since H4 and H5 are similar welds.
H5 VISUAL INSPECTIONS: Inspection of "punchmarked cracks" for length on ID. UT INSPECTIONS: Reinspection of two (2) areas inspected during RFO B109R1 to determine crack growth. H6a VISUAL INSPECTIONS: None scheduled. and H6b LJT INSPECTIONS: Inspect three (3) accessible areas between jet pumps. Inspections to be performed from the OD to look for ID-l connected cracking. I H7 No inspections scheduled. Inspection tools / techniques being developed by BWRVIP. H8 No inspections scheduled. inspection tools / techniques being developed by BWRVIP. H9 VISUAL INSPECTIONS: None scEeduled. UT INSPECTIONS: 100% of circumference scheduled. V1 - V11 VISUAL INSPECTIONS: None scheduled. The allowable axial flaw size determined by l References 6 and 7 is greater than the widest plate. l. Therefore, the vertical flaws are bounded by analysis. UT INSPECTIONS: None scheduled. Shroud Support No inspections scheduled inspection tools / techniques being developed by BWRVIP. Legs Repair VISUAL INSPECTIONS: Inspect two (2) clamps for general appearance, missing parts, Clamps and integrity of tack welds. UT INSPECTIONS: Not Applicable
l . Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 2 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 15 of 27
, Table 4 provides a summary of the RFO B110R1 inspection findings. Reference 12 3
contains the detailed Inservice Inspection results. 1 TABLE 44 UNIT 1: RFO.B110R1 DETAILED INSPECTION RESULTSj UWELO4 4 RESULTSS i H1 VISUAL INSPECTIONS: None performed. UT INSPECTIONS: Four (4) areas were examined: between Shroud Head Bolt Lug Sets 3-4; 14-15; 26-27; and 33-34. A total of eight (8) OD surface connected planar flaws were detected and maximum observed depth was 0.728". One ID surface connected planar flaw was detected and maximum observed depth was 0.354". Compared to previous sizing data,it was concluded that no change in size occurred during Fuel Cycle #9. H5 VISUAL INSPECTIONS:
- Punch marked cracks" on ID were inspected for length. No changes from RFO B109R1 were noted (i.e. no crack growth).
UT INSPECTIONS: Two (2) areas inspected during RFO B109R1 were reinspected to determine crack growth. The examination was performed from the inside surface and from below the weld. Compared to previous sizing data, it was concluded that no change in size occurred during Fuel Cycle #9. H6a VISUAL INSPECTIONS: None performed. UT INSPECTIONS: Three (3) areas were examined: between the Jet Pumps @ 75.5' azimuth; @ 225.5' azimuth; and @ 315.5* azimuth. A total of two (2) planar flaw type indications were detected, however only one was determined to be surface connected. The maximum observed depth of the ID connected indication was 0.354". H6b VISUAL INSPECTIONS: None performed. UT INSPECTIONS: Three (3) areas were examined: between the Jet Pumps @ 75.5* azimuth; @ 225.5' azimuth; and @ 315.5' azimuth. A total of three (3) planar flaw type indications were detected, however only two were determined to be surface connected (both were ID surface connected). The maximum observed depth of the ID connected indications was 0.551" H9 VISUAL INSPECTIONS: None performed. UT INSPECTIONS: 100% of circumference was inspected from the RPV OD. No indications were detected. l Repair VISUAL INSPECTIONS: Inspection of two (2) clamps for general appearance, missing parts, Clamps and integrity of tack welds was satisfactory.
. - - - ._ - - - . . . .. . - =
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~ Carolina Power & Light Company ESR # B5-00765
' Brunswick Nuclear Plant .
Revision 1 ' A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 16 of 27 ( ANALYSIS i t '. The RFO B110R1 indications were evaluated by one of the following methods: l l 1)- Comparison of crack depth and length data with conditions assumed in l } previous analyses (References 14 and 15). I i I !
- 2) Analysis for structural significance by screening indications in accordance with . l l Reference 5, as supplemented by References 6 and 7.
l ,
- I
- 3) Performing weld-specific structural analysis to determine the allowable crack '
l size in accordance with Reference 13. l l I i The screening process (method 2) includes several conservatisms, such as assuming i : that all cracks are through-wall. It then provides a bounding crack length for initial l screening. Cumulative effective crack lengths which are smaller than the bounding l -i crack length are not a structural concern and are considered acceptable. l I- i Flawed welds can be specifically analyzed (method 3) by the "BWR Core Shroud l l Distributed Ligament Length (DLL) Computer Program" (Reference 13.) This program l
~
was prepared for the BWRVIP Assessment Subcommittee by GE to evaluate the l structural margins for a given set of flaw c6nfigurations in the shroud for both the l l upset and faulted loading conditions. H6b was evaluated using the DLL analysis i method since inspection coverage did not allow- application of the (method 2) l screening criteria. ' l I The H6b analysis bounds the H6a condition since: (1) the applied stresses are higher l at H6b than at H6a and (2) the maximum recorded flaw depth is greater at H6b than j at H6a. Analysis conservatively assumed a 360* flaw with the maximum recorded l flaw depth and a crack growth rate of 0.1" per cycle through the next three fuel l cycles. (Attachment C provides a summary of loads at each horizontal weld location l and was used in conjunction with Reference 13.) The 0.1" per cycle growth rate was l established by UT equipment uncertainties, as described in Attachment E, and is j considered conservative based on comparison of B110R1 inspection results to l B109R1 inspection results. Attachment A presents the results of the H6b analysis. l The analysis results are summarized in Table 6. t l i Table 5 provides a summary of the analysis results for each weld joint. [
I . Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 1 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 17 of 27 1 6
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TABLE 5fUNIT CANALYSIStAND RESULTS1SUMMARVA 4%
' ^
4WFI n o . # ANAL YSIUAND RFEUl_ TEE NNI H1 Four (4) areas were examined that had been previously examined in RFO B109R1. A total of nine planar flaw type indications were detected. i
- 1) Comparison to RFO B109R1 data is shown in Attachment D, along with l postulated crack depths at the end of two additional fuel cycles (RFO l B112R1) and a bounding curve from Reference 15. The postulated extent of l cracking in RFO B11281 is fully bounded by Reference 15, except for the l two indications listed below, which were not addressed by the analysis. l l
- 2) Indication #2 between lugs 26 and 27 was not previously reported by GE in l the RFO B109R1 report, however a review performed by GE (Reference 10) l demonstrates that the same indication was located in RFO B109R1. l Connection to the OD surface was indeterminate by GE since a 45' shear l wave transducer was not used on the lower side of H1, l l
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- 3) Indication #1 between lugs 3 and 4 was not previously reported by GE in the l RFO B109R1 report, however a review (Reference 16) demonstrates that the l same indication was located in RFO B109R1, but was interpreted as l geometry. l
-~
l _
- 4) Reference 15 does not specifically evaluate the two indications above. The j indications are both below the H1 weld and therefore the depth of the H1 l reinforcement fillet weld leg cannot be added to the remaining ligament for l structural evaluation. However, the remaining net section area evaluated in l Reference 15 is less than that projected for the section immediately below l the H1 weld at the end of the next two fuel cycles. Therefore, the Reference l 15 evaluation fully bounds the identified condition below the H1 weld. A l detailed discussion is included in Attachment E. l O
I Carolina Power & Light Company ESR # 95-00765 , Brunswick Nuclear Plant Revision 1 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 18 of 27 de EThBLE!5tf0N'ITf1? ANAL SISTANDlRESULTSiSOMMARYAN l w sa. ' . , .
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44% % ,, _2 ,,N ' WWELD n <+ 'd,ANALYSIR' AND RFSULTS ' > > % ' l H5 Two (2) areas inspected during RFO B109R1 were reinspected to determine crack l growth. Cracking below the weld in the 169* and 274* areas was detected by VT and sized by UT during RFO B109R1. l 1 l
- 1) The RFO B110R1 UT inspections revealed no changes in crack depth or l l
length, as depicted in Attachment D, and therefore the condition of H5 is l l fully bounded by previous arialysis. l l I
- 2) A deviation in reported flaw locations from RFO B109R1 (GE) and RFO l 8110R1 (Siemens) was evaluated by GE and concluded to be a reporting l error in the 8109R1 report. Appropriate correction demonstrates close l correlation between B109R1 and B110R1 results. l H6a Three (3) areas were examined: between the Jet Pumps @ 75.5* azimuth: @ 225.5*
azimuth; and @ 315.5* azimuth. A total of two (2) planar flaw type indications were and detected in H6a and three (3) in H6b. 1 i H6b 1) A structural iritegrity analysis of the core shroud was performed for both the l upset and the faulted loading conditions and is included as Attachment A. l H6b was analyzed as the limiting case since the applied stresses are higher at l H6b than at H6a. The resultyf the Attachment A analyses are summarized l in Table 6. l I l
- 2) The analyses conclude that the core shroud is structurally adequate for l continued operation during Fuel Cycles 10 and 11. l H9 UT examination of 100% of the circumference identified no reportable indications.
Repair Inspection of two (2) clamps (spanning H2 and H3) for general appearance, missing Clamps parts, and integrity of tack welds was satisfactory. L
. =
Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 1 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 19 of 27 c iTABLE 61
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- H68 SAFETv F cTORs FoR WAu.TNickNEss:1 1 M,!E[
~w LusiNo NRC CRACK GROWTH VALUES 3 [m' ~ ig (1) Required (2) (3) REMAINING SAFETY LOADING Safety WELD CRACK GROWTH ASSUMED WALL FACTOR l CONDITION Factor DESIGNATION ALLOWANCE FLAW SIZE THICKNESS l (in.) (in.) (in.)
NORMAL / UPSET 2.77 H6b 0.6 0.551 0.349 5.71 FAULTED 1.39 H6b 0.6 0.551 0.349 3.53 H6B SAFETY factors FoR WALL THICKNESS UstNO BNP U1 CALCULATED CRACK GROWTH VALUES (1) Required (4) (3) REMAINING SAFETY LOADING Safety WELD CRACK GROWTH ASSUMED WALL FACTOR l CON 0 MON Factor DEstGNATION ALLOWANCE FLAW SIZE THICKNESS l _ (in.) (in.) (in.) NORMAL / UPSET 2.77 H6b 0.1 0.551 0.849 13.72 (1 cycle)
- ,. FAULTED 1.39 H6b 0.1 , 0.551 0.849 8.31 (1 cycle) '
NORMAL / UPSET 2.77 H6b 0.3 0.551 0.649 10.52 (3 cycles) FAULTED 1.39 H6b 0.3 0.551 0.649 6.39 (3 cycles) l l (1) includes Power Uprate pressures and asymmetric loading for the shroud. (2) NRC mandated allowance of 5x104 in/hr for one cycle of operation = 0.6*/ cycle for BNP. (3) Flaw size used is maximum observed crack depth for the H6b weld and is assumed 360* thru-wall. (4) The crack growth used for BNP U1 is 0.1*/ cycle and is based on the results of the H1 weld inspections.
l . 1 Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 1 A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 20 of 27 SijNIMARY 4 The BNP-1 Refueling Outage (RFO) 8110R1 Core Shroud inspections are complete and evaluated in this ESR. Core shroud relevant indications from the RFO B110R1 inspections have been evaluated by methods 1,2 or 3 as presented in the previous , section and have been judged to be acceptable for continued operation of BNP-1 for two cycles of operation. This ESR concludes that structural integrity of the core shroud will be maintained, l with full FSAR safety margins, for at least the next two operating cycles based on analysis of the inspection results. Crack growth experienced during Cycle 9 was substantially less than postulated by previous analysis. The inspection results from 8110R1 show that the existing l condition is essentially unchanged from the condition identified in B109R1. 1 i Furthermore, the postulated crack lengths at the end of Cycle 11, based upon l B110R1 results, are fully bounded by previous analyses and will not reduce the structural design margins below allowable values. Therefore, the condition of the core shroud does not impose any restrictions to BNP-1 operation during the next two i operating cycles. The BNP-1 Core Shroud is " acceptable as is" for Operating Cycles j 10 and 11. l i I Future inspection plans will consider not only these inspection results, but will also consider continuing developments in the industry, to ensure utilization of the best information and technology to address the issue. REFERENCES
- 1. FP-50096, Sheet 1 of 2, " Assembly and Finish Machining Shroud Core Structure," Revision 2
- 2. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section lil,1965 Edition through Summer,1967 Addenda
- 3. EER 93-0536, Evaluation of Unit 1 Core Shroud Indications and Operability I Assessment of Unit 1 and 2, Revision 1
( l
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I.. . . . Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 2 A TTACHMENT TO ENGINEERING SERWCE REQUEST Page 21 of 27
- 4. GE Report NEDC 32300-P, " Brunswick Unit 1 Shroud Sample Metallurgical d Evaluations," dated October 1993.
- 5. GE Report GENE-523-123-0993, Rev. 2, " Evaluation and Screening Criteria for the Brunswick 1 Shroud Indications," November 1993.
- 6. RAM-94-092/ SIR-94-029, " Addendum to the Brunswick Unit 1 Screening Criteria", dated 4/6/94.
,l-
- 7. RAM-94-099/ SIR-94-031, " Minimum Required Unflawed Core Shroud Material at Brunswick, Units 1 and 2", dated 4/11/94.
- 8. GE report R129, RFO B109R1 Data Sheets from OPT-90.1, in-Vessel Visual l Examination !
i 9. System Description SD-01, " Nuclear Boiler," Revision 26, dated Nov.1,1993.
- 10. Memorandum, A. R. Jaschke (GE) to John Langdon (CP&L), " Review of Unit 1 1993 Shroud Weld H1 UT Data", dated May 5,1995
- 11. Memorandum, A. R. Jaschke (GE) to John Langdon (CP&L), "1993 and 1995 '
Core Shroud Weld H5 Data for Unit 1", dated May 10,1995 l 12. RFO B110R1 IVVI andtlT reports, including OPT-90.1 data sheets and video l cassettes.
- 13. GE Report GENE-523-113-0894, Supplement 1, "BWR Core Shroud Distributed Ligament Length Computer Program"
- 14. CP&L Calculation No. 1-821-0049, Revision 0
- 15. GE Report GENE-523-144-1093, " Analysis of Unit 1 Welds H1, H2, & H3...",
Revision 1, November,1993
- 16. Memorandum, E. Black (CP&L) to W. B. Wilton (CP&L), " Review of H1 Weld UT Indication at 30 Degree Azimuth", dated April 17,1995 DOCUMENT UPDATES No document updates are required as a result of this ESR. !
l 1 ESR ACTION ITEMS No ESR action items are required as a result of this ESR. Future inspections and , reportings are governed by OPT-90.1. l
1 .
- l Carolina Power & Light Company Brunswick Nuclear Plant '
ESR # 95-00765 Revision 1 \ A TTACHMENT TO ENGINEERING SERVICE REQUEST Page 22 of 27 l l I I c _e l 5 _
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l l i ESR # 95-00765 i Attachment A Revision 0 l Page 1 of 18 1 l H6B Wald Location: Assumed Crack Growth = 0.6"/ cycle: Upset Load case- 1 l DLL: DISTRIBUTED LIGAMENT LENGTH EVALUATION (REVISION: ' 10/07/94) DATE OF CURRENT ANALYSIS: 05/14/1995 :
SUMMARY
OF INPUTS: Angle increment .= 1.0 dag. (COARSE)
. Membrane Stress, Pa = 309.. psi '
Bending Stress, Pb = 2375. psi ; Safety Factor, SF = 2.77 Mean Radius, Ra = 88.75 inches i Wall Thickness, t = 1.500 inches 1 Material = i 1 304 SS ; Stress Intensity, Sa = 16900. psi Fluence = 1.9E+19 n/cm*2 (Thus, LEFM evaluation not applicable) THETA 1 THETA 2 THICKNESS , REGION [deg.) (deg.) [ inches) ! 1 .0 360.0 .349 t !
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LIMIT LOAD RESULTS: i i ALPHA MOMENT Pb8 SAFETY i (deg) (in-lbs) [ psi) FACTOR RESULT l
\
.0 5.569E+08 150 05~. 5.71 ACCEPTABLE -
5.0 5.569E+08 15005. 5.71 ACCEPTABLE ' 10.0 5.569E+08 15005. 5.71 ACCEPTABLE 15.0 5.569E+08 15005. 5.71 ACCEPTABLE 20.0 5.569E+08 15005. 5.71 ACCEPTABLE 25.0 5.569E+08 15005. 5.71 ACCEPTABLE i 30.0 5.569E+08 15005. 5.71 ACCEPTABLE 35.0 5.569E+08 15005. 5.71 ACCEPTABLE ! 40.0 5.569E+08 15005. 5.71 ACCEPTABLE 45.0 5.569E+08 15005. 5.71 ACCEPTABLE , 50.0 5.569E+08 15005. 5.71 ACCEPTABLE 55.0 5.569E+08 15005. 5.71 ACCEPTABLE 60.0 5.569E+08 15005. 5.71 ACCEPTABLE 65.0 5.569E+08 15005. 5.71 ACCEPTABLE , 70.0 5.569E+08 15005. 5.71 ACCEPTABLE 75.0 5.569E+08 15005.
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5.71 ACCEPTABLE 80.0 5.569E+08 15005. 5.71 ACCEPTABLE
il-.. .. ESR # 95-00765 Attachment A Revision 0 Page 2 of 18 85.0 5.569E+08 15005. 5.71 90.0 ACCEPTABLE ' 5.569E+08 15005. 5.71 ACCEPTABLE 95.0 5.569E+08 15005. 5.71-100.0 ACCEPTABLE 5.569E+08 15005. 5.71 ACCEPTABLE 105.0 5.569E+08 15005. 5.71 110.0 ACCEPTABLE
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5.569E+08 15005. 5.71 ACCEPTABLE 115.0 5.569E+08 15005. 5.71 120.0 ACCEPTABLE 5.569E+08 15005. 5.71 ACCEPTABLE 125.0 5.569E+08 15005. 5.71 130.0 ACCEPTABLE 5.569E+08 15005. 5.71 ACCEPTABLE 135.0 5.569E+08 150'05. 5.71 140.0 ACCEPTABLE 5.569E+08 15005. 5.71 ACCEPTABLE-145.0 5.569E+0S 15005. 5.71-150.0 ACCEPTABLE
' 5.569E+08 15005. 5.71 ACCEPTABLE 155.0 5.569E+08 15005. 5.71 160.0 ACCEPTABLE 5.569E+08 15005. 5.71 ACCEPTABLE 165.0 5.569E+08 15005. 5.71 170.0 ACCEPTABLE 5.569E+08 15005. 5.71 ACCEPTABLE 175.0 5.569E+08 15005. 5.71 180.0 ACCEPTABLE 5.569E+08 15005. 5.71 ACCEPTABLE 185.0 5.569E+08 15005. 5.71 ACCEPTABLE 190.0 5.569E+08 15005, 5.71 195.0 ACCEPTABLE 5.569E+08 15005. 5.71 ACCEPTABLE 200.0 5.569E+08 15005, 5.71 205.0 ACCEPTABLE
'~ 5.569E+08 15005. 5.71 ACCEPTABLE 210.0 5.569E+08 15005, 5.71 ACCEPTABLE 215.0 5.569E+08 15005. 5.71 ACCEPTABLE 220.0 5.569E+08 15005. 5.71 ACCEPTABLE 225.0 5.569E+08 15005. 5.71 ACCEPTABLE 230.0 5.569E+08 15005. 5.71 ACCEPTABLE
,. 235.0 5.569E+08 15005, 5.71 ACCEPTABLE 240.0 5.569E+08 150U5. 5.71 ACCEPTABLE -
245.0 5.569E+08 15005. 5.71 ACCEPTABLE 250.0 5.569E+08 15005, 5.71 ACCEPTABLE 255.0 5.569E+08 15005, 5.71 ACCEPTABLE 260.0 5.569E+08 15005. 5.71 ACCEPTABLE 265.0 5.569E+08 15005. 5.71 ACCEPTABLE 270.0 5.569E+08 15005. 5.71 ACCEPTABLE 275.0 5.569E+08 15005. 5.71 ACCEPTABLE 280.0 5.569E+08 15005. 5.71 ACCEPTABLE 285.0 5.569E+08 15005. 5.71 ACCEPTABLE 290.0 5.569E+08 15005. 5.71 ACCEPTABLE 295.0 5.569E+08 15005. " 71 ACCEPTABLE 300.0 5.569E+08 15005. 5.71 ACCEPTABLE 305.0 5.569E+08 15005. 5.71 ACCEPTABLE 310.0 5.569E+08 15005, 5.71 ACCEPTABLE 315.0 5.569E+08 15005. 5.71 ACCEPTABLE L 320.0 5.569E+08 15005. 5.71 ACCEPTABLE
l . . ESR # 95-00765 Attachment A Revision 0 Page 3 of 18 325.0 5.569E+08 15005. 5.71 ACCEPTABLE 330.0 5.569E+08 15005. 5.71 ACCEPTABLE 335.0 5.569E+08 15005. 5.71 ACCEPTABLE 340.0 5.569E+08 15005. 5.71 ACCEPTABLE 345.0 5.569E+08 15005. 5.71 ACCEPTABLE g 350.0 5.569E+08 15005. 5.71 ACCEPTABLE. 355.0 5.569E+08 15005. 5.71 ACCEPTABLE ACCEPTABLE! MINIMUM SAFETY FACTOR = 5.71 AT 155.0 DEGREES. o e t he
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I. - I J ESR # 95-00765 i A,ttachment A Revision 0 l Page 4 of 18 H6B Wald Location: Assumed Crack Growth = 0.6"/ cycle: Faulted Load Case l DLL: DISTRIBUTED LIGAMENT LENGTH EVALUATION (REVISION: 10/07/94) i 1 DATE OF CURRENT ANALYSIS: 05/14/1995 I
SUMMARY
OF INPUTS: Angle increment = 1.0 dag. (COARSE) Membrane Stress, Pa = 1099. psi Banding Stress, Pb = 3423. psi : Safety Factor, SF = 1.39 Mean Radius, Ra = 88.75 inches Wall Thickness, t = 1.500 inches Material = 304 SS Stress Intensity, Sa = 16900. psi Fluence = 1.9E+19 n/ca"2 (Thus, LEFM evaluation not applicable) THETA 1 THETA 2 THICKNESS I REGION [deg.] (deg.) (inches) 1 .0 360.0 .349
~~
-L LIMIT LOAD RESULTS:
ALPHA MOMENT Pb' SAFETY
,, [deg) (in-lbs) (psi) FACTOR RESULT
.0 5.513E+08 14854. 3.53 ACCEPTABLE 5.0 5.513E+08 14854. 3.53 ACCEPTABLE 10.0 5.513E+08 14854. 3.53 15.0 ACCEPTABLE 5.513E+08 14854. 3.53 ACCEPTABLE 20.0 5.513E+08 14854. 3.53 ACCEPTABLE 25.0 5.513E+08 14854. 3.53 ACCEPTABLE 30.0 5.513E+08 14854. 3.53 ACCEPTABLE 35.0 5.513E+08 14854. 3.53 ACCEPTABLE :
40.0 5.513E+08 14854. 3.53 ACCEPTABLE 45.0 5.513E+08 14854. 3.53 ACCEPTABLE 50.0 5.513E+08 14854. 3.53 ACCEPTABLE 55.0 5.513E+08 14854. 3.53 ACCEPTABLE 60.0 5.513E+08 14854. 3.53 ACCEPTABLE 65.0 5.513E+08 14854. 3.53 ACCEPTABLE 70.0 5.513E+08 14854. 3.53 ACCEPTABLE o 75.0 5.513E+08 14854. 3.53 ACCEPTABLE
- , - - ., , - ,_ - = -n-
a S ESR # 95-00765 Attachment A Revision 0 : Page 5 of 18 80.0 5.513E+08 14854. 3.53 ACCEPTABLE 85.0 5.513E+08 14854. 3.53 ACCEPTABLE 90.0 5.513E+08 14854. 3.53 ACCEPTABLE 95.0 5.513E+08 14854. 3.53 ACCEPTABLE 100.0 5.513E+08 14854. 3.53 ACCEPTABLE 105.0 5.513E+08 14854. 3.53 ACCEPTABLE 3
.110.0 5.513E+08 14854. 3.53 ACCEPTABLE
- 115.0 5.513E+08 14854. 3.53 ACCEPTABLE ;
120.0 5.513E+08 14854. 3.53 ACCEPTABLE
, 125.0 5.513E+08 14854. 3.53 ACCEPTABLE '
130.0 5.513E+08 14854. 3.53 ACCEPTABLE 135.0 5.513E+08 14854. 3.53 ACCEPTABLE 140.0 5.513E+08 14854. 3.53 ACCEPTABLE i 145.0 5.513E+08 14854. 3.53 ACCEPTABLE f 150.0 5.513E+08 14854. 3.53 ACCEPTABLE ' 155.0 5.513E+08 14854. 3.53 , ACCEPTABLE 160.0 5.513E+08 14854. 3.53 ' ACCEPTABLE ! 165.0 5.513E+08 14854. 3.53 ACCEPTABLE 170.0 5.513E+08 14854. 3.53 ACCEPTABLE : 175.0 5.513E+08 '14854. 3.53 ACCEPTABLE ! 180.0 5.513E+08 14854. '3.53 ACCEPTABLE ! 185.0 5.513E+08 14854. 3.53 ACCEPTABLE ! 190.0 5.513E+08 14854. 3.53 ACCEPTABLE l 195.0 5.513E+08 14854. 3.53 ACCEPTABLE
'~
200.0 5.513E+08 14854. 3.53 ACCEPTABLE 205.0 5.513E+08 14854.* 3.53 ACCEPTABLE 210.0 5.513E+08 14854. 3.53 ACCEPTABLE 215.0 5.513E+08 14854. 3.53 ACCEPTABLE 220.0 5.513E+08 14854. 3.53 ACCEPTABLE 225.0 5.513E+08 14854. 3.53 ACCEPTABLE 230.0 l'
.. 5.513E+08 14854. 3.53 ACCEPTABLE 235.0 5.513E+08 14854. 3.53 ACCEPTABLE -
240.0 5.513E+08 14854. 3.53 ACCEPTABLE 245.0 5.513E+08 14854. 3.53 ACCEPTABLE 250.0 5.513E+08 14854. 3.53 ACCEPTABLE i 255.0 5.513E+08 14854. 3.53 ACCEPTABLE 260.0 5.513E+08 14854. 3.53 ACCEPTABLE 265.0 5.513E+08 14854. 3.53 ACCEPTABLE 270.0 5.513E+08 14854. 3.53 ACCEPTABLE 275.0 5.513E+08 14854. 3.53 ACCEPTABLE 280.0 5.513E+08 14854. 3.53 ACCEPTABLE 285.0 5.513E+08 14854. 3.53 ACCEPTABLE 290.0 5.513E+08 14854. 3.53 ACCEPTABLE 295.0 5.513E+08 14854. 3.53 ACCEPTABLE 300.0 5.513E+08 14854. 3.53 ACCEPTABLE 305.0 5.513E+08 14854. 3.53 ACCEPTABLE ! 310.0 5.513E+08 14854, 3.53 ACCEPTABLE { t 315.0 5.513E+08 14854. 3.53 ACCEPTABLE I I
- l. . * * '
ESR # 95-00765 Attachment A Revision 0 Page 6 of 18 i i 320.0 5.513E+08 14854. 3.53 325.0 ACCEPTABLE 5.513E+08 14854. 3.53 ACCEPTABLE 330.0 5.513E+08 14854. 3.53 335.0 ACCEPTABLE 5.513E+08 14854. 3.53 ACCEPTABLE 340.0 5.513E+08 14854. 3.53 345.0 ACCEPTABLE 5.513E+08 14854. 3.53 ACCEPTABLE 350.0 5.513E+08 14854. 3.53 355.0 ACCEPTABLE 5.513E+08 14854. 3.53 ACCEPTABLE ACCEPTABLE! MINIMUM SAFETY FAdrOR = 3.53 AT 155.0 DEGREES.
;w O
O .
~ I.' . . i i ESR # 95-00765 A,ttachment A- Revision 0 Page 7 of 18 H6B Wald Location: Assumed Crack Growth = 0.1"/ cycle: 1 cycle of growth: Remaining Wall Thickness = 0.849": Upset Load Case DLL: DISTRIBUTED LIGAMENT LENGTH EVALUATION (REVISION: 10/07/94)
- l. DATE OF CURRENT ANALYSIS: 05/14/1995
SUMMARY
OF INPUTS: Angle increment = 1.0 deg. (COARSE) Membrane Stress, Pa = 309. psi , Bending Stress, Pb = 2375. psi ' Safety Factor, SF = 2.77 Mean Radius, Ra = 88.75 inches j
; Wall Thickness, t = 1.500 inches Material = 304 SS Stress Intensity, Sm = 16900. psi Fluence = 1.9E+19 n/cm*2 (Thus, LEFM evaluation not applicable)
THETA 1 THETA 2 THICKNESS REGION [deg.] [deg.] (inches) 1 1 .0 360.0 .849 1
'~
LIMIT LOAD RESULTS: ALPHA MOMENT Pb' SAFETY
,, (deg] (in-lbs) (psi) FACTOR RESULT t
I
.0 1.356E+09 36524. 13.72 ACCEPTABLE .
5.0 1.356E+09 36524. 13.72 ACCEPTABLE ! 10.0 1.356E+09 36524. 13.72 ACCEPTABLE 15.0 1.356E+09 36524. 13.72 ACCEPTABLE 20.0 1.356E+09 36524. 13.72 ACCEPTABLE 25.0 1.356E+09 36524. 13.72 ACCEPTABLE ; 30.0 1.356E+09 36524. 13.72 ACCEPTABLE ! 35.0 1.356E+09 36524. 13.72 ACCEPTABLE < 40.0 1.356E+09 36524. 13.72 ACCEPTABLE 45.0 1.356E+09 36524. 13.72 ACCEPTABLE 50.0 1.356E+09 36524. 13.72 ACCEPTABLE 55.0 1.356E+09 36524. 13.72 ACCEPTABLE 60.0 1.356E+09 36524. 13.72 ACCEPTABLE 65.0 1.356E+09 36524. 13.72 ACCEPTABLE 70.0 1.356E+09 36524. 13.72 ACCEPTABLE o 75.0 1.356E+09 36524. 13.72 ACCEPTABLE ; i O
-i , . - - - , . _ , , , - , , .-
il..- . j ESR # 95-00765 i Attachment A Revision 0 l Page 8 of 18 i i r 80.0 1.356E+09 36524. 13.72 ACCEPTABLE I 85.0 1.356E+09 36524. 13.72 ' ACCEPTABLE 90.0 1.356E+09 36524. 13.72 ACCEPTABLE 95.0 1.356E+09 36524. 13.72 ACCEPTABLE 100.0 1.356E+09 36524. 13.72 ACCEPTABLE I
' 105.0 1.356E+09 36524. 13.72 ACCEPTABLE j 110.0 1.356E+09 36524. 13.72 ACCEPTABLE 115.0 1.356E+09 36524. 13.72 i ACCEPTABLE 120.0 1.356E+09 36524. 13.72 ACCEPTABLE 125.0 1.356E+09 36524. 13.72
~
ACCEPTABLE 130.0 1.356E+09 3652'4. 13.72 ACCEPTABLE 135.0 1.356E+09 36524. 13.72 ACCEPTABLE 140.0 1.356E+09 36524. 13.72 ACCEPTABLE 145.0 1.356E+09 36524. 13.72 ACCEPTABLE 150.0 1.356E+09 36524. 13.72 ACCEPTABLE 155.0 1.356E+09 36524. 13.72 ACCEPTABLE 160.0 1.356E+09 36524. 13.72 ACCEPTABLE 165.0 1.356E+09 ' 36524. 13.72 170.0 ACCEPTABLE 1.356E+09 36524. 13.72 ACCEPTABLE 175.0 1.356E+09 36524. 13.72 ACCEPTABLE 180.0 1.356E+09 36524. 13.72 ACCEPTABLE 185.0 1.356E+09 36524. 13.72 ACCEPTABLE 190.0 1.356E+09 36524. 13.72 ACCEPTABLE 195.0 1.356E+09 36524. 13.72 200.0 ACCEPTABLE
*~
1.356E+09 36524. 13.72 ACCEPTABLE 205.0 1.356E+09 36524i- 13.72 210.0 ACCEPTABLE 1.356E+09 36524. 13.72 ACCEPTABLE 215.0 1.356E+09 36524. 13.72 ACCEPTABLE 220.0 1.356E+09 36524. 13.72 225.0 ACCEPTABLE 1.356E+09 36524. 13.72 ACCEPTABLE 230.0 1.356E+09 36524. 13.72
~
ACCEPTABLE 235.0 1.356E+09 365244 13.72 ACCEPTABLE - 240.0 1.356E+09 36524. 13.72 ACCEPTABLE' 245.0 1.356E+09 36524, 13.72 250.0 ACCEPTABLE 1.356E+09 36524. 13.72 ACCEPTABLE 255.0 1.356E+09 36524. 13.72 ACCEPTABLE 260.0 1,356E+09 36524. 13.72 ACCEPTABLE 265.0 1.356E+09 36524. 13.72 270.0 ACCEPTABLE 1.356E+09 36524. 13.72 ACCEPTABLE 275.0 1.356E+09 36524. 13.72 ACCEPTABLE 280.0 1.356E+09 36524. 13.72 ACCEPTABLE 285.0 1.356E+09 36524. 13.72 ACCEPTABLE 290.0 1.356E+09 36524. 13.72 ACCEPTABLE 295.0 1.356E+09 36524. 13.72 ACCEPTABLE 300.0 1.356E+09 36524. 13.72 1 ACCEPTABLE l 305.0 1.356E+09 36524. 13.72 ACCEPTABLE 310.0 1.356E+09 36524. 13.72 i ACCEPTABLE l e 315.0 1.356E+09 36524. 13.72 ACCEPTABLE 1 i
i ESR # 95-00765 e a n0 A,ttachment A Page 9 of 18 320.0 1.356E+09 36524. 13.72 ACCEPTABLE 325.0 1.356E+09 36524. 13.72 ACCEPTABLE 330.0 1.356E+09 36524. 13.72 ACCEPTABLE 335.0 1.356E+09 36524. 13.72 ACCEPTABLE i 340.0 1.356E+09 36524, 13.72 ACCEPTABLE 345.0 1.356E+09 36524. 13.72 ACCEPTABLE 350.0 1.356E+09 36524. 13.72 ACCEPTABLE 355.0 1.356E+09 36524. 13.72 ACCEPTABLE ACCEPTABLE! MINIMUM SAFETY FACTOR = 13.72 AT .0 DEGREES. D 1
-e , ,
~~ .
i l l 6, l k
ESR # 95-00765
, Attachment A Revision 0 l' Page 10 of 18 H6B Wald Location: Assumed Crack Growth = 0.1"/ cycle: 1 cycle
'of grwoth: Remaining Wall Thickness = 0.849": Faulted Load Case DLL: i DISTRIBUTED LIGAMENT LENGTH EVALUATION (REVISION:
10/07/94) y DATE OF CURRENT ANALYSIS: 05/14/1995
SUMMARY
OF INPUTS:
. Angle increment = 1,0 deg. (COARSE)
Membrane Stress, Pm = 1099. psi Bending Stress, Pb = 3423. psi ! Safety Factor, SF = 1.39 Mean Radius, 2n = 88.75 inches i Wall Thickness, t = 1.500 inches Material = 304 SS Stress Intensity, Sa = 16900. psi Fluence = 1.9E+19 n/ca"2 (Thus, LEFM evaluation not applicable) THETA 1 THETA 2 THICKNESS REGION [deg.] [deg.] [ inches) 1 .0 360.0 .849 LIMIT LOAD RESULTS: ALPHA MOMENT Pb' SAFETY
.- [deg) [in-lbs) [ psi] FACTOR
----- ...------- _ RESULT ;
-------------- 1
. 0 1.354E+09 36468. 8.31 ACCEPTABLE j 5.0 1.354E+09 36468. 8.31 10.0 ACCEPTABLE !
1.354E+09 36468. 8.31 ACCEPTABLE 15.0 1.354E+09 36468. 8.31 20.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 25.0 1.354E+09 36468. 8.31 30.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 35.0 1.354E+09 36468. 8.31 40.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 45.0 1.354E+09 36468. 8.31 50.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 55.0 1.354E+09 36468. 8.31 60.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 65.0 1.354E+09 36468. 8.31 70.0 ACCEPTABLE A 1.354E+09 36468. 8.31 ACCEPTABLE 75.0 1.354E+09 36468. 8.31 ACCEPTABLE
,+9- . - . - y , - . - . ~ . , -
ESR # 95-00765 Attachment'A Revision 0 Page 11 of 18 i 80.0 1.354E+09 36468. 8.31 ! 85.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 90.0 1.354E+09 36468. 8.31 95.0 ACCEPTABLE : 1.354E+09 36468. 8.31 ACCEPTABLE 100.0 1.354E+09 36468. 8.31 105.0 ACCEPTABLE ! 1.354E+09 36468. 8.31 ACCEPTABLE 110.0 1.354E+09 36468. 8.31 115.0 ACCEPTABLE 1.354E+09 36468.- 8.31 ACCEPTABLE 120.0 1.354E+09 36468. 8.31 ACCEPTABLE 125.0 1.354E+09 36468. 8.31 130.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 135.0 1.354E+09 36468. 8.31 140.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 145.0 1.354E+09 36468. 8.31 150.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 155.0 1.354E+09 36468. 8.31 160.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 165.0 1.354E+09 36468. 8.31 170.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 175.0 1.354E+09 36468. 8.31 180.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 185.0 1.354E+09 36468. 8.31 190.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 195.0 1.354E+09 36468. 8.31 200.0 ACCEPTABLE
~ 1.354E+09- 36468. 8.31 ACCEPTABLE 205.0 1.354E+09 364657 8.31 210.0 ACCEPTABLE 1.354E+09 36468. - 8.31 ACCEPTABLE 215.0 1.354E+09 36468. 8.31 220.0 ACCEPTABLE 1.354E+09 36468.- 8.31 ACCEPTABLE 225.0 1.354E+09 36468. 8.31 230.0 ACCEPTABLE 1.354E+09 36468_. 8.31 ACCEPTABLE 235.0 1.354E+09 36468. 8.31 240.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 245.0 1.354E+09 36468. 8.31-250.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 255.0 1.354E+09 36468. 8.31 260.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 265.0 1.354E+09 36468. 8.31 270.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 275.0 1.354E+09 36468. 8.31 280.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 285.0 1.354E+09 36468. . 8.31.
' 290.0 ACCEPTABLE-1.354E+09 36468. 8.31 ACCEPTABLE 295.0 1.354E+09 36468. 8.31 300.0 ACCEPTABLE
- 1.354E+09 36468. 8.31 ACCEPTABLE 305.0 1.354E+09 36468. 8.31 310.0 ACCEPTABLE 1.354E+09 36468. 8.31 ACCEPTABLE 315.0 1.354E+09 36468. 8.31 ACCEPTABLE m -- ..y. . , - . - -
.,l-1 v
ESR # 95-00765 g Attachment A Revision 0 s Page 12 of 18 320.0 1.354E+09 36468. 8.31 ACCEPTABLE 325.0 1.354E+09 36468. 8.31 ACCEPTABLE ! 330.0 1.354E+09 36468. 8.31 ACCEPTABLE i 335.0 1.354E+09 36468. 8.31 ACCEPTABLE 340.0 1.354E+09 36468. 8.31 ACCEPTABLE t 345.0 1.354E+09 36468. 8.31 ACCEPTABLE 350.0 1.354E+09 36468. 8.31 ACCEPTABLE - 355.0 1.354E+09 36468. 8.31 ACCEPTABLE ACCEPTABLE 1 MINIMUM SAFETY FACTOR = 8.31 AT 350.0 DEGREES. ! I i- ) i i l'
~
g. j l l l e m 0
. , . , . - , . - ,. .w
I'. . <
\
l ESR # 95-00765
, Attachment A Revision 0 i Page 13 of 18 i
M6B Wald Location: Assumed Crack Growth = 0.1"/ cycle: of growth: 3 cycles Remaining Wall Thickness = 0.649: Upset Load Case i DLL: DISTRIBUTED LIGAMENT LENGTH EVALUATION (REVISION: 10/07/94) g DATE OF CURRENT ANALYSIS: 05/14/1995 SUMP'ARY OF INPUTS:
. Angle increment =
1,0 deg. (COARSE) Membrane Stress, Pm = 309. psi Bending Stress, Pb = 2375. psi Safety Factor, SF = 2.77 Mean Radius, Rm = 88.75 inches 1 i Wall Thickness, t = 1.500 inches Material = 304 SS ' Stress Intensity, Sm = 16900. psi Fluence = 1.9E+19 n/cm"2 (Thus, LEFM evaluation not applicable) THETA 1 THETA 2 THICKNESS REGION [deg.) [deg.) [ inches) t 1 .0 360.0 .649 E-LIMIT LOAD RESULTS:
===========-
t ALPHA MOMENT Pb' i SAFETY
,, (deg) (in-lbs) (psi,) FACTOR
---------- RESULT
.0 1.036E+09 27920.
10.52 ACCEPTABLE 5.0 1.036E+09 27920. 10.52 ACCEPTABLE 10.0 1.036E+09 27920. 10.52 ACCEPTABLE 15.0 1.036E+09 27920. 20.0 10.52 ACCEPTABLE 1.036E+09 27920. 10.52 ACCEPTABLE 25.0 1.036E+09 27920. 10.52 ACCEPTABLE 30.0 1.036E+09 27920. 10.52 35.0 ACCEPTABLE 1.036E+09 27920. 10.52 ACCEPTABLE 40.0 1.036E+09 27920. 10.52 45.0 ACCEPTABLE 1.036E+09 27920. 10.52 ACCEPTABLE 50.0 1.036E+09 27920. 10.52 55.0 ACCEPTABLE 1.036E+09 27920. 10.52 ACCEPTABLE 60.0 1.036E+09 27920. 10.52 65.0. ACCEPTABLE 1.036E+09 27920. 10.52 ACCEPTABLE 70.0 1.036E+09 27920. 10.52 o 75.0 ACCEPTABLE 1.036E+09 27920. 10.52 ACCEPTABLE
E b, -
- i i
1 ESR # 95-00765 Attachment A Revision 0 Page 14 of 18 80.0 1.036E+09 27920. . 10.52 ACCEPTABLE 85.0 1.036E+09 27920. 10.52 ACCEPTABLE-90.0 1.036E+09 27920. 10.52 95.0 ACCEPTABLE 1.036E+09 27920. 10.52 ACCEPTABLE 100.0 1.036E+09 27920. 10.52 ACCEPTABLE
- )- 105.0 1.036E+09 27920. l'O.52 ACCEPTABLE 110.0 1.036E+09 27920. 10.52 ACCEPTABLE 115.0 1.036E+09 27920. 10.52 ACCEPTABLE 120.0 1.036E+09 27920. 10.52 ACCEPTABLE 125.0 1.036E+09 27930. 10.52 ACCEPTABLE 130.0 1.036E+09 27920. 10.52 ACCEPTABLE 135.0 1.036E+09 27920. 10.52 ACCEPTABLE 140.0 1.036E+09 27920. 10.52 ACCEPTABLE
' 145.0 1.036E+09 27920. 10.52 ACCEPTABLE 150.0 1.036E+09 27920. 10.52 ACCEPTABLE 155.0 1.036E+09 27920. 10.52 ACCEPTABLE 160.0 1.036E+09 27920. 10.52 ACCEPTABLE 165.0 1.036E+09 27920. 10.52 ACCEPTABLE 170.0 1.036E+09 27920. 10.52 ACCEPTABLE 175.0 1.036E+09 27920. 10.52 ACCEPTABLE 180.0 1.036E+09 27920. 10.52 ACCEPTABLE 185.0 1.036E+09 27920. 10.52 ACCEPTABLE 190.0 1.036E+09 27920. 10.52 ACCEPTABLE 195.0 1.036E+09 27920. 10.52 ACCEPTABLE
'~ 200.0 1.036E+09 27920. 10.52 ACCEPTABLE 205.0 1.036E+09 279207 10.52 ACCEPTABLE 210.0 1.036E+09 27920. 10.52 ACCEPTABLE 215.0 1.036E+09 27920. 10.52 ACCEPTABLE 220.0 1.036E+09- 27920. 10.52 ACCEPTABLE 225.0 1.036E+09 27920. 10.52 ACCEPTABLE
,, 230.0 1.016E+09 27920. 10.52 ACCEPTABLE 235.0 1.036E+09 27926. 10.52 ACCEPTABLE ~
I 240.0 1.036E+09 27920, 10.52 ACCEPTABLE 245.0 1.036E+09 27920. 10.52. ACCEPTABLE 250.0 1.036E+09 27920, 10.52 ACCEPTABLE 255.0 1.036E+09 27920. 10.52 ACCEPTABLE 260.0 1.036E+09 27920, 10.52 ACCEPTABLE 265.0 1.036E+09 27920. 10.52 ACCEPTABLE 270.0 1.036E+09 27920. 10.52 ACCEPTABLE 275.0 1.036E+09 27920. 10.52 ACCEPTABLE 280.0 1.036E+09 27920. 10.52 ACCEPTABLE 285.0 1.036E+09 27920, 10.52 ACCEPTABLE 290.0 1.036E+09 27920. 10.52 ACCEPTABLE 295.0 1.036E+09 27920. 10.52 ACCEPTABLE ! 300.0 1.036E+09 27920. 10.52 ACCEPTABLE ! 305.0 1.036E+09 27920. 10.52 ACCEPTABLE 310.0 { 1.036E+09 27920. 10.52 ACCEPTABLE 315.0 1.036E+09 27920. 10.52 ACCEPTABLE l l
- f,. , .
= >
ESR # 95-00765 Attachment A Revision 0 Page 15 of 18 320.0 1.036E+09 27920. 10.52 ACCEPTABLE - 325.0 1.036E+09 27920. 10.52 ACCEPTABLE 330.0 1.036E+09 27920. 10.52 ACCEPTABLE 335.0 1.036E+09 27920, 10.52 ACCEPTABLE 340.0 1.036E+09 27920. 10.52 ACCEPTABLE-T' 345.0 1.036E+09 27920. 10.52 ACCEPTABLE 350.0 1.036E+09 27920. 10.52 ACCEPTABLE 355.0 1.036E+09 27920. 10.52 ACCEPTABLE ACCEPTABLE! MINIMUM SAFETY. FACTOR = 10.52 AT 25.0 DEGREES. , m I i i 1 i m
h.c. . . . ESR # 95-00765
- Attachment A Revision 0 Page 16 of 18 H6B Wald Location: Assumed Crack Growth = 0.1"/ cycle: 3 cycles of growth: Remaining Wall Thickness = 0.649":
Faulted Load Case 10/07/94) DLL: DISTRIBUTED LIGAMENT LENGTH EVALUATION (REVISION: ' DATE OF CURRENT ANALYSIS: 05/14/1995
SUMMARY
OF INPUTS: i
, Angle increment =
1.0 deg. (COARSE) Membrane Stress, Pm = 1099. psi Bending Stress, Pb = 3423. psi Safety Factor, SF = 1.39 Mean Radius, Rm = 88.75 inches Wall Thickness, t = 1.500 inches Material = 304 SS , Stress Intensity, Sm = 16900. psi ' Fluence = 1.9E+19 n/cm"2 (Thus, LEFM evaluation not applicable) THETA 1 THETA 2 THICKNESS REGION [deg.) (deg.) [ inches) 1 .0 360.0 .649
~
.A LIMIT LOAD RESULTS:
_========= ALPHA MOMENT Pb' SAFETY I
, [deg)
---~~
(in-lbs) (pai) FACTOR RESULT ,
.0 1.032E+09 27801. 6.39 ACCEPTABLE 5.0 1.032E+09 27801. 6.39 10.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 15.0 1.032E+09 27801. 6.39 20.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 25.0 1.032E+09 27801. 6.39 30.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 35.0 1.032E+09 27801. 6.39 40.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 45.0 1.032E+09 27801. 6.39 50.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 55.0 1.032E+09 27801. 6.39 60.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 65.0 1.032E+09 27801. 6.39 70.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE t 75.0 1.032E+09 27801. 6.39 ACCEPTABLE v-~ w, - , - - , v w ~ -
l.d . .. .
.I ESR # 95-00765 Attachment A Revision 0 Page 17 of 18 80.0 1.032E+09 27801. 6.39 85.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 90.0 1.032E+09 27801. 6.39 ACCEPTABLE 95.0 1.032E+09 27801. 6.39 100.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE
,, 105.0 1.032E+09 27801. 6.39 it- 110.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 115.0 1.032E+09 27801. 6.39 120.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE
, 125.0 1.032E+09 27801. 6.39 130.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE l
135.0 1.032E+09 27801. 6.39 ACCEPTABLE 140.0 1.032E+09 27801. 6.39 145.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE L 150.0 h 1.032E+09 27801. 6.39 ACCEPTABLE 155.0 1.032E+09 27801. 6.39 160.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 165.0 1.032E+09 27801. 6.39 170.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 175.0 1.032E+09 27801. 6.39 180.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 185.0 1.032E+09 27801. 6.39 190.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 195.0 1.032E+09 27801. 6.39
,200.0 ACCEPTABLE
'" 1.032E+09 27801. 6.39 ACCEPTABLE 205.0 1.032E+09 278017 6.39 210.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 215.0 1.032E+09 27801. 6.39 220.0 ACCEPTABLE '
1.032E+09 27801. 6.39 ACCEPTABLE 225.0 1.032E+09 27801. 6.39 230.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 235.0 1.032E+09 2780f. 6.39 ~ 240.0 ACCEPTABLE-1.032E+09 27801. 6.39 ACCEPTABLE 245.0 1.032E+09 27801. 6.39 250.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 255.0 1.032E+09 27801. 6.39 260.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 265.0 1.032E+09 27801. 6.39 270.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 275.0 1.032E+09 27801. 6.39 280.0 ACCEPTABLE 1.032E+09 27801. 6.39 ACCEPTABLE 285.0 1.032E+09 27801. 6.39 290.0 ACCEPTABLE 1.032E+09 27801. 6.39 ' ACCEPTABLE 295.0 1.032E+09 27801. 6.39 ACCEPTABLE 300.0 1.032E+09 27801. 6.39 i ACCEPTABLE
~
305.0 1.032E+09 27801. 6.39 ACCEPTABLE 310.0 1.032E+09 27801. 6.39 i A ACCEPTABLE 315.0 1.032E+09 27801. 6.39 ACCEPTABLE {
. . ~ . . ,-- -. _ , _ . ._ _ _
ll . . ESR # 95-00765 ? Attachment A Revision 0 Page 18 of'18 320.0 1.032E+09 27801. 6.39
' ACCEPTABLE t 325.0 1.032E+09 27801. 6.39 ACCEPTABLE 330.0 1.032E+09 27801. 6.39 ACCEPTABLE I 335.0 1.032E+09 27801. 6.39 ACCEPTABLE 340.0 1.032E+09 27801. 6.39 ACCEPTABLE
- y 345.0 1.032E+09 27801. 6.39 ACCEPTABLE ;
350.0 1.032E+09 27801. 6.39 ACCEPTABLE I 355.0 1.032E+09 27801. 6.39 ACCEPTABLE ACCEPTABLE! MINIMUM SAFETY FAdTOR = 6.39 AT 25.0 DEGREES. l J
~
;_ .I
_- 1 a t
i e., . =,.- i ! l l i ESR 95-00765 Attachment B Page 1 of 5 1 Safety Review Cover Sheet (with Signatures) .i j Page On File ' 5 1 i
]
1 r l
y . ESR 95-00765 Att:chment D Page 2 of 5 Revision 1
, 10 CFR50.59 Propam Manual Rev. 3 i 1'
ATTACHMENT A - CP&L SAFETY REVIEW PACKAGE.
' PART 1: SAFETY ANALYSIS (See instructions in Section 8.4.1) ;
(Attach additional sheets as necessary.) ' 1 DOCUMENT NO. n non1M REV.NO._L i DESCRIPTION OF CHANGE: Invessel Visual inspection (IVVI) of selected Core Shroud welds was performed per OPT-90.1 - during the B110R1 outage. Additionally, Ultrasonic UT) inspections were performed on selected welds. Inspections revealed minimal growth in cracks which had been identified and analyzed during the previous Refueling Outage (RFO B109R1). Relevant indications were identified in welds H6s and H6b in areas that had not been previously UT inspected. ANALYSIS: The reactor internals perform the following safety related design basis functions as i specified in the UFSAR: 1 1
- 1. Provide a floodable volume in which the core can be adequately cooled in the event of r
a breach in the nuclear system processJaarrier external to the reactor vessel.
- 2. Umit deflections and deformation to assure that the control rods and the core standby cooling systems can perform their safety functions during abnormal operational
)
transients and accidents. _1
- 3. Assure that the safety design bases (1) and 12) above are satisfied so that the safe shutdown of the plant and removal of decay heat are not impaired.
Interpanular stress corrosion cracking DGSCC) of the type and form experienced with recirculation piping and related systems in Bolling Water Reactors (8WRs)is the cause of cracking. Crack extension is possibly assisted by neutron fluence and " oxide wedging" at certain locations. Susceptible material conditions, Ngh residual stress from fabrication, and exposure to a strong oxidizing environment are - sufficient to produce the cracking observed. Because these factors are not consistently present across l
~
the shroud, the location and degree of cracking varies across the shroud. 4.
- - - - -,-,y-. - --
e ESR 95-00765 Attachment B Page 3 of 5 Revision 1 The core shroud must maintain a floodable volume above the two-thirds core height elevation. The q
;. cracks are caused by intergranular stress corrosion cracking, and inherently are tight. Any through wall 8
cracks would result in negligible leakage into the downcomer region and the leakage would be contained by the reactor presswa vessel. The Emergency Core Cooling systems provide sufficient make-up and cooling capacity to answo that the fuel will remain covered. EER 93-0536 was issued to assess Unit 1 shroud structural integrity after RFO B109R1 and to justify i. continued operation for one cycle. 'Ihe RFO B110R1 inspections are complete and ESR 95-00765 provides results of the analysis of the cracking on the Unit 1 shroud, and therefore serves as an update to EER 93-0536, Revision 1. Structuralintegrity of the core shroud will be maintained, with fug FSAR safety margins, for a minimum of two operating cycles based on analysis of the inspections , performed. 'i The current inspection results show no significant changes in crack length or patterns, except for the
- new indications identified by UT on welds H6a and H6b. A weld-specific structural analysis i
(Attachment A to ESR 95-00765) was performed for these indications in accordance with ASME Section III, Appendix A guidelines and results were acceptable. , Estimated crack lengths and pattoms through the end of Fuel Cycle #11 (two fuel cycles from the time i t of this evaluation) are fully bounded by the previous analyses (References 1,2 and 31, based on _ trending from inspection results and on conservative analyses. Therefore, aN conclusions reached in the analyses remain valid for the next two fuel cycles. The Core Shroud cracks have been previously ; t addressed in a 10 CFR50.59 Safety Evaluation in EER 934536, Revision 1 (Reference 4). Since the ' Safety Evaluation considered crack growth rates and end-of cycle crack lengths, pattoms, and
-:- i consequences postulated by the analyses, the SafetyEvaluation fully bounds the current condition and '
l i supports continued operation for the next two fuel cycles. ;
REFERENCES:
11 General Electric Report # GE-NE-523-123-0993, Revision 2, Noveraber,1993
- 2) Structural Integrity Report # RAM-94-092/$1R-94-029, April,1994
- 3) Structural integrity Report # RAM-94-099/ SIR-94-031, April,1994 !
- 4) EER No. 93-0536 and associated 10 CFR50.59 Safety Evaluation, Revision 1 A
e m -
i ll_ 2 .. . - ESR 95-00765 Attachment B Page 4 of 5 Revision 1 10 CFR50.59 Program Manual Rev. 3 ATTACHMENT A
- i CP&L SAFETY REVIEW PACKAGE PART 11: ITEM CLASSIFICATION DOCUMENT NO. RF.on n1Md REV.NO. 1 Yas bio
- 1. Does this item represent:
, a. A change to the facility as described in the SAFETY ANALYSIS REPORT 7
[] [XI
- b. A change to the procedures as described in the SAFETY ANALYSIS [] [XI REPORT 7
- c. A test or experiment not described iri the SAFETY ANALYSIS REPORT 7 [] [XI
- 2. Does this item involve a change to the individual plant Operating License or to [1 [X1 its Technical Specifications?
- 3. Does this item require a revision to the FSAR7
[] [X)
- 4. Does this item involve a change to the Off-Site Dose Calculation Manual? [] [XI !
- 5. Does this item constitute a change to the Process Control Program? [] [XI 6.
j Does this item involve a major change to a Radweste Treatment System? [] [XI i
- 7. Does this item involve a change to the Technical Specification Equipment List (BSEP and SHNPP only)? [] [XI
- 8. Does this item impact the NPDES Permit (all 3 sites) or constitute an [] [XI
*unreviewed environmental question" (SHNPP Environmental Plan, Section 3.11
'or a "significant environmental impact" (BSEPl?.-.
- 9. Does this item involve a change to a previously accepted:
- n. Quality Assurance Program ,
- b. [] IXl 4 Security Plan (including Training, Qualification, and Contingency Plans)? [] [XI
- c. Emergency Plan 7
" d. Independent Spent Fuel Storage installation license? (if *yes," refer to Il [X]
Section 2.4.2, " Question 9," for speci&l considerations. Complete Part [] [X) VIin accordance with Section 8.4.6) SEE SECTION 8.4.2 FOR INSTRUCTIONS FOR EACH "YES* ANSWER. REFERENCES. List FSAR and Technical Specification references used to answer questions 19 above. Identify specific reference sections used for any "Yes" answer. UFRAR Raetinne 1_9 9 K 11_191 9 109K1 t o R 1_ K M_1 K_2 1 1_? 1 M d_i_ 7 1 1 1_M K-7.7_1_1 9 9 741 0149 anaf PF.2.r.:c r i K iinie 1 Techniemi Cameirner Ir_. R;.re!. . .: 2/d 1 1 anaf 2 r -!=.02.d F_ !
- . . - - ~ ~ .--.- .
- 3. g _.
g , t' !
't ESR 95-00765 Attachment 8 !
Page 5 of 5 ! Revision 1 l ]
' ATTACHMENT 2 i
10 CFR50.59 Program Manual Rev. 3
^
i' PART Ill: UNREVIEWED SAFETY QUESTION DETERMINATION SCREEN i .j- DOCUMENT NO. sp an-n1'id REV.NO. 1
~
i i Y.as hin .
.- . 1. Is this change fully addressed by another completed IX]- [] !
UNREVIEWED SAFETY QUESTION determination? (See Sections 7.2.1, 7.2.2.5, and 7.9.1.1) REFERENCE DOCUMENT: FFR hin QLOR1R REV.NO. 1
- 2. Yan- hin For procedures, is the change a non-intent change which only (check all that apply): (See Section 7.2.2.3) -i
-INA] (NAl
[] - Corrects typographical errors which do not alter ' the meaning or intent of the procedure; or, 11 Adds or revises steps for clarification (provided they are consistent with the original purpose or ! applicability of the procedure); or,
~
11 Changes the title of an organizatiolial position; or, il Changes names, addresses, or telephone numbers of persons; or, i il Changes the designation of an item of equipment where the ,
" equipment is the same as the original equipment or is an authorized replacement; or, ;.
[] Changes a specified tool or instrument to an equivalent substitute; or, [] Changes the format of a procedure without altering the meaning, intent, or content; or il Deletes a part or all of a procedure, the deleted portions of which are wholly covered by approved plant procedures? If the answer to either Question 1 or Question 2 in PART lit is "Yes,' then PART IV need not be completed. O y - emgr- m -.+m,,--e.. - ,ym - , ws,- -*w+r . -- -,--v - w-- .-- - m.-,-----w -
I ? I . { - c Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 1 A TTACHMENT C Page 1 of 6
?.............. .. .
a). -
References:
BWR. . Core shrou..,d Evaluedon *Leen Definition GuideEne'. SL 4942. Rev.0 : Syrt> ole:
~~
i D = Deed I. cede .. . This sproedef.eetuses me obsolute summetod for conebedon of stessee
' 8 =Bu'oishe9 Forces ' . d5d maynAnd to'bi'erkAlined So seesmodate hisnt specise requiremen(s
' ~ ~
P =. I.nteme.l P. roseuro. Differendet
. ; . for the cor.rt>inedon'ofloede and steesee. '
; Pn =NormelPressurel '
( PU =dpeet Proseure S ! ~ : Pt =Feutted Pressure +
- MSLOCA'= Mein Steem LOCAL
, ; RRLOCA Neeetor Reclrisuladon LOCA OBE =0peratog Besie Eerthquake 4 g SSE ISdfo Shuidhwn E$requake ;
L Pm =PMierf Moivbrans gtpossee c : O SP. b. = Prima,ry8,5N., n.~. ding'. Swe' eees ; f }I
- Inputet : '0.0535g'
~ ' '" T..i .s:t .o +0 . : 10 Pel ?
1030700 . . V!T u att p y Q . . (11'.8 Poi}
-: 0.036 b. fin *3 !! W,5 :I : ' t A.,
- 29.4 Peif'
~ '
, 0.290 b.lin $ L'$
E
~ ?fE t' kE *b f 20 Pol [
927317.915 tn.*2i:. O n.... 11 s:r / 22.4 Phi 5-60.000 KipsL ' m~s *: 1 A1 4 13i Peli em.=x !0.195 K/lb.. F R I + + i c. t :: i . .(177l du_ N JCJF ' .s 10.875 in. ,
.ve - .
. ex
[ 7474.8184. e 2 lrE. *;1,1
. - . . . .n.... ...,. ... .. .
Notes: 1; The sheded heedings suprosent the design inputs used b calculete the stesses..
)% P4ssiriin shnery menta6rsund binside stresses ism vos K enioid e'elienevos
- 43. NAgative OSEse fErloeds irWicate met $4Y 56:UpwErd opposien gravity.Y i
F4/. ?^' o ..Nsgs, si v.ssd.i.forseesseeindiceu tinsion7^ ^ ' ? 3
~
-p
SUMMARY
OF LOADS AND STRESSES AT EACH HORIZONTAL WELD LOCATION - Brunswick Unit I 4 .
~ ~ .a
.. .. _ _._ J i i 1
( . e Carolina Power & Light Company ESR.# 95-00765 Brunswick Nuclear Plant Revision 1 A TTACHMENT C Page 2 of 6
~'
e m w J shisu'd ? ,i L Shroud d J.'f tP s . . c rr )
'1 W t '
if 'lan4{- 'shreve:L cs nt,'rt:4. : h,GD, @ R R l.O CA 51Cf) OV1 J! $ Wall! SSkEtion ' '!!T> - '10 l'L '1 !Mokwnt lW4NI* ' E 46:n l i d"I ]l h ;rri DnMJ > f d'rds ' N64ush(i.. ;? h t .n - # 1rii Il ?- ' 3 44Eeusds DeelOEtien ~ a it ' i i 1 i '
, t lib IIh.h INN.Y2f '!!dd83 fit .itg:YO; ' Ir . ;_ (' _ _
,] e lin Kip'el:
ML ' EN + J750q 3.250 g, 1.500fg . 929 : . ^ 4.164E104 g LO10E 44:.' 2.650E104j 7.313E 4 1 2.130E 402-
!. H25 356.963 i .750 it 3.250 t I.500 ' .929T -.
4.164E 404 ; 2.830E 44' 3.460E 44 "- J.841E 42 2 6.039E 42-N3h 353.963} 8.750( l 7d50) 1.500[ 29380.f ; 3.650Eg . 2 530ENj' 3.480Ek){ j.290E.402l 6.508E 42 H4: 317.9635: .750$ 7.250 ; I.500 ;i 29.3P0;
- N8) 3.650E 44 !.: - 3.670E 44 -? 4.890E 44i 9.676E 42 : 1.350E 43 219.7131 S.750 m^ 7.250 I.500) ?? 29.380 it;L i 3.650E 404 / 4 5.620E M ! B.880E 404l 1.596E 403 6
' 4.541E 431 EHAMj l 185.7i3j 8.750I 7.250 ) 25380J , 5.650Eh ' T.890E44)f' 1.000E M i.4S9E 43; 1.606) l5.180E43 i H68? -
179.4633: 5.500 T 4.0001 1.500 1 98.750 R 3.385E 44s B.040E 44!: : 1.080E.45 -a f.800E 431 6.390E 43:
. H;7,2
- 127.400.3 y
- 5. 500. Y. 3. 500 I .
2.000 "> 06.1.858, 5 ' 4387E N...! 3.970E 404 ?. 1.350E405 1.498E 44. s 9351.E103 1
SUMMARY
OF LOADS AND STRESSES AT EACH HORIZONTAL WELD LOCATION - Brunswick Unit 1
, - _ _ , 2 u __ -.- _ .-
i , I {- . a. Carolina Power & Light Company ESR R 95-00765 Brunswick Nuclear Plant . ' Revision 1 A TTACHMENT C Page 3 of 6 - l
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i .',~...'..s .. .
. .,~.....y r , , . ..
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. Ve rtic al b E ffe ctiva* e E ffe ctive y ' Be nding . f aending '.
I ShtEbIf lb i- 3.1 I t . Buoyahti: N:bS'kf ' l'S $ E N ' '! Neiglit :/ :INhidEk [OBEN
'sSE?
YWeld$' =) 1
- i. i1 O EForeAY [tJplift! :. IIpfifth ?UBET[ l's'sEl .[. : Pil. "P b' DAeIEnEti$n E{ U h O ' INhi)I [I'KipeN^ 0(KIpAIS ~l lKbl [lKipel
- i IP sil1 5 (PElli
. H1; 19140; . rang _nn:; q : 126.003 15.64) ,6.74 :.- 13.48 j 103.62.; D6A8; . 535.35
,- 442.69 ; s
- H2 ;
25em0 + 12m0 ; - 137.10; 17o2:( . 7.33 ? 14.67.:. ti2.7s '. t05.41 : s31.58 -830.91 1H3; l 254 00 1 12.00i i48.60-.- 18.4s - 7.ss 1 Is.90 - : 122.20 ? 114.2s 7' r26.64 : '948.06
.: H4$44. 286.00 j ' 2.06 $ 1 1U7.80 19.59! $.44Il 16.85[. 129.77;; 1005.61 < F 1339.90 TH51 - 129.00 ] >
.00 E _'
121.33) 182.90[ 140.55 - ' 22.89 ... 9.78 ': 19.56 < + 150.33: 1813.93 ? 2433.18 M6d$ 41d.0d[ - 192.06.' 23A3 Id.i73 20.54i 157.89 1 1k7.62 $: f H68 ! - Sid.00]j ' 28.00{i 28.00) : Zi3.50 i 2161 92 { 2904N8
)H7~
26.54 j , 11.4d f 22.88$ . 175.85y i 15438) 2375.21 { . 3190.58 (14.003 28.001 227.10 ;' 28.19? 12.15 t 24.30 ^ ' 186.762 174.61f 2221.9si :::3008.71 n H8E 41dAOb < 28D0. h 227.10 ; 28dS.i ,. $h2.$,5i " -
- ] 54 30L. $ 186j6 5 v
174.61. Y 2340.1 2- ' 3164.71 n -
SUMMARY
OF LOADS AND STRESSES AT EACH HORIZONTAL WELD LOCATION - Brunswick Unit I t
_ 7 I , I { . r . Carolina Power & Light Company Brunswick Nuclear Plant ESR # 95-00765 . A TTACHMENT C Revision 1 Page 4 of 6
^
e
- Upeet K MS LOCA RRLOCA ftRLOCA:
Ve rtic al Ve rtical N ormat F a ulte d ' B e nding Be nding _ _ . Deed Load Buoyency 08E' ~SSEi Pre ssure Co mbine d Combine d S hroud
~
Pre ssure S tresod e - Stresses Oombin e d
' S t'rd ese e ? S tr'e s'o e s" Stre ssee S tre s s e e S tre sse s Stre sse e Stroese e . Stressee - Slovedovvn Ac oustic Stre sse s Weld ' . j! P m ' 'Pm ~Pm* .. Pm' Pm L D 48 +P u 40 8 E D 48 4P u 408 E D e elsn e tfor
'Pm' ' 'Pb3 ' P h* O 48 #n
. [P oll IPell . IPoll { , IPell , IP oll . ' IP ell ' (Pell
' H1 - IP ell IP ell- IPed ' IP sil Co mp re s elon
. f 42.224 .-11.e55 -I.505 --35.218 Tenelon
- H2- 154.753 +19.211
- )-305.J5J .-306.555 -1.156 5.115 -153.755 - 231.M ~B7T.583
-8.279 16.559 308.353 906.559 H3 179.170 1 9.224 .14.503 -172.811 450.494
-22.242 '-9.586 19.171 -329,377 -812.675
-968.369 11.755 ,17.833 -172.449 N4 ?190.263 :.*23.619 f 10.179 do.358 538.604 -902.674 i'329.377 -968.369 26.513 36.986 162.734 L 145 .220.405 27.361 11.792 23.583 832.694 1178.519
. -329.377 . 968.369 -125.922 124.422
- H6A: 231.498 1-28.738 112.385 -24.770
' -136.333 1865.803 1982.052 I
-509.628
~
H68 0267.668 : -1256.771 204.379 ' $69.344 13065 68 -1842.664 L -33.228 14.320 -28.641 -529.171 -2481.170
' -j 304.965 232.192 188.789 ;294.731
'H7. L 213A70 -26.549 -11.442 -22.884 2066.156 -2684.258
-398.053 -981.620 333250 l 206.163 H8- - 213.870 -26.549 11.442 -22.884 210.732 1999.811 2444.158 3
.-398.053 -981.620 333.850 . 227.411 l
-210.732 2117.930 -2562.278
SUMMARY
OF LOADS AND STRESSES AT EACH HORIZONTAL WELD LOCATION - Brunswick Unit I
I i 1 ( . e - Ccrofina Power & Light Company Brunswick Nuclear Plant ESR # 95-00765 , A TTACHMENT C Revition 1 Page 5 of 6 wry , , Ec.n*en.*n iconsin e.. q co nsin.e .
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- c.nein.d 's ir. ...'. ' sir. .... . p 4 *i4sE o 4 erase o 4 #t4esE o4 #fasE
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i o 4 *t4ssE o'd #t4sE l RR I.0cA I ' R R l.OCAi f RR LOCA ' " R R iOCA " Shr.ud - .D4 #f5 0 4 *P A4ssE D 4MesE !Ms 1.oca T ' MaiOCA5 .slAwd.wn BIAwd.wn - "As.v.'Ne 'Ae.uisis jNdidY Ms$1.0c'Al ' !(Psil [ [ [P.'il f M .' (P.li f os.isk.s.n "tP.i:V ' c.ner...l.n- 0 Tinsi.n . 6.#r... .n : T.sii.4 : c.w r... .n 1 7.n41. : c.mp a..i.n f(P.ll" [IPillf
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-m.wso w.m. T. n.s.n n- , ym-u .-i m.m us.wo *u.es
. N2 n p[ ,-771D17 645.536 '1020.275 an.iw -
. -s.o. sos U 43.330 FjiSja.481 '650.700 -j029.499 858.039 1034.779
. HS -l M11.441 r 758.444 k1139.885 -117.452 4 1778.677 .768.199 -1151A40 - 774.277 '1157.Si8
[N4J ?-801.726 1156.803 mis 22.907 f 517.811 , j216i.979 - 1183.316
- 15AS.500 1993.790 : 1559.973 H56 , 775J25 s
-1834.276
[2273.268 . a2593.100 1 4 232.092 y2399.190 ; -2719D22 - ; 2397.690 - 2717.522
'H6AT 41054.010 i2572.838 D 3236.114 11825.896 T-3983.257 12777.217 M4440.~493 52742.182 Neg l- 'i1070.525 .-3405.458
~U2867.205 E-3513.948 [209I.4'12 1 4 289.742 3099.397 [ 3748,1'40 s .;3055.095
/ -794.299 ' -3702.737 H7 -? t 12775.090 > -3242.321 !2191.523 :-3s25. set . o $10s.940 "
;H8. :-357s.172 29s1.253 L-344s.4s4 I794.299 [2931.096 3398.328 d 2347.530
-390j.895 (3264.947 /d732.i?9 {3158.507 ?-3625.739
SUMMARY
OF LOADS AND STRESSES AT EACH HORIZONTAL WELD LOCATION - Brunswick Unit 1
1 a' t - Ccrofina Power & Light Company Brunswick Nuclear Plant ESR.# 95-00765 ' Revision 1 A TTACHMENT C Page 6 of 6
- ce mameer.: comermes ' . .comm me s :-- . commines y .ccommenee -- e comnines , r cemninedg '
P rima ry' f P rima ry - - P rime ry P rimary y _ Primary ) : Primary-
' P rima ry ?
Me mbrane i J Bending - Me mbrene J B e ndin8 : : Membr'ene : . Be'ndin8 Sendina
'S tre s ee d S tre ssee ? - stre sse s' Sireesee [' [ S tre sse s" - L S t'rk eeee'- .' S tre sse e - E mer8 enc y E me r8ency Shrewd - - Uheef " Uheek [ Faulte'd MS ~ Favited - Feultel'RR' Fasited
^ . F o ulte d t' 3 21 YWsNb O 45 4Pd 40sEv : osEN ' D+54Pt4ssEv :ssEV D4e4PtassEv psEk4SIAddwn s$'Eh '4Aheuelle D 4 4Pn 4sSE D 4e 4Ms LOCA Deel8NetieE ilPellP** JIPoll I ' ' " {IP 5Il ' II[All
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(Nil ? " IP5ll " H1 . -is3.ase . -(si.20s . quas.as : -iss. coa H2. q181.080
. . a= 3 == .
aas.!< < -esi.so sai.as - -is i.ss 6631.58 = 787.575 '-83031 -189.370 l -840.13 841.54 1.-845.41
-771.02 93 c:-182.035 ;-720.64 3 830.613 :-948.06 -151.620 -95922 4 66.90 :756.44
-811.44 H 4 : 172.913 '.1005.61 (-822.084 -1339.90 183.082 1366.41 61376.88 l 1156.80 801.73 Hs .~ -148.124 c-1813.93 -798.908 2433.18 159.916 r +2559.11 2557.61 J 2273.27
, L 315.253 -775.32
- HSA; g2161.92 ,.72904.48 -331.638
'.31078.700 ., ?. 3108.86 ~!-3073.82 Has - J309.051 { 1099.165 - .73190.58 l72572.84 $054.01
}2375.21 ---323.371 -3422.77 D -3379.37
- .. 2867.21 1070.52
'H7 l -222.174 : -2221.98 -817.183 .--3008.71 233.816 ' -3342E6 -3214.87 2775.09
-794.30
;HS-f222.174 .'21 340.10 -817.183 - h3164.71 -233.616
- jd498.56 ; 794.30 o
- -3392.12 .i2931.10
.w - w -
SUMMARY
OF LOADS AND STRESSES AT EACH HORIZONTAL WELD LOCATION - Brunswick Unit 1 _ _ _ _ _ _ _ _ _ . _ __ ___ _ _ _ _ _ _ _ . _ . -- - . . , _ - - _ _ - - - ~ . . - _ - - - - _-
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.-. l . - Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 1 A TTACHMENTE Pa0e 1 of 6 i DISCUSSION OF INDICATIONS BELOW WELD HI BACKGROUND The H1 weld was inspected by visual and ultrasonic methods during Refuel Outage (RFO) B109R1, as summarized in Table 1 below. b
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$_hy',+ w4TABLEi1E UNITS 1LWELD;H 17 tNSPECTION ! RESULTS:tRFO!B109R1 E E M mefg M WhisTg;Es%gamesn g g usMm qgsa g w ees%98 gfg % Wgg VISUAL INSPECTIONS: 100% of OD and ID surfaces were inspected. The cracks are long but not continuous. Primary orientation is circumferential, located on the OD, mainly below the bolting lugs in the Separator Support Ring. Approximately 268' of the circumference (74%) contains cracks. No consistent cracking pattom exists, except for some branching associated with attachment welds. Cracks have been found only on the outer surface.
UT INSPECTIONS: Measuremer:ts were made staight (8) locations. Depths ranged from < 0.3* to 0.7*. 1 m The H1 indications did not meet the preliminary screening criteria of Reference 1, as documented by References 2 and 3. Therefore, specific analysis (Reference 4) was performed by General Electric (GE) in accordance with ASME Section XI
)
(Reference 5) techniques for flawed austenitic stainless steel weldments. The analysis considered postulated "end of cycle" (EOC) crack depths and assumed 360' crack length. l _ The analysis results demonstrated that the flawed H1 weld had sufficient ' structural margin to justify continued operation of BNP-1. In fact, calculated I l
1 . . l l l Carolina Power & Light Company ESR # 95-00765 ; Briinswick Nuclear Plant Revision 1 l A TTACHMENTE Page 2 of 6 l l safety factors were more than twenty (20) times the required safety factors. A summary is shown in Table 2 below. i l I f g q+y + ' - wW n,*e s .e.2.x*':.wwwhe:.rwwemm wumesme,s+wm ww3w.>
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g g M-.mepWjWjiDEP1)gjg [gNggS?J !y 1 i l Upset inh _EI _I _h ' 2.25 1.03 1.22 1.06 ksi 47.7 2.25 1 l Faulted 2.25 1.03 1.22 1.93 kai 26.3 1.125 During RFO B110R1, the H1 weld was UT examined in four areas ( 30', 60* , 140', and 260') that had been previously examined during RFO B109R1. Table 3 provides a summary of the RFO B110R1 inspection findings.
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__$ WELD!H1t MT@f M&!W$46Me2&dRSMd;lNSPECTIONiftESULTSltRFO'B11 14Ss9EASEW23MAM461%Eh$$Mni . . , VISUAL INSPECTIONS: None performed. UT INSPECTIONS: Four (4) areas were examined: between Shroud Head Bolt Lug Sets 3-4: 14-15; 26-27; and 33-34. A total of eight (8) OD surface connected planar flaws were detected and maximum observed depth was 0.728". One ID surface connected planar flaw was detected and maximum observed depth was 0.354". L I
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e j i Carolina P6wer & Ught Company ESR N 95-00765 Brunswick Nuclear Plant Revision 1 A TTACHMENTE .i Pace 3 of 6 ANALYSIS t I l Four (4) areas were UT examined that had been previously examined in RFO ' B109R1. A total of nine planar flaw type indications was detected. ! is M-Comparison to RFO B109R1 data is shown in Attachment D to ESR #95-00765, ! along with postulated crack depths at the end of two additional fuel cycles (RFO ; B112R1) and a bounding curve from Reference 4. The postulated extent of cracking in RFO B112R1 is explicitly bounded by Reference 4, except for the two l j i indications listed below, which were not addressed by the analysis. I
- 1) Indication #2 between lugs 26 and 27 was not previously reported by GE in the RFO B109R1 report, however a review performed by GE (Reference 6) demonstrates that the same indication was located in RFO B109R1. It was j not reported because its connection to the OD surface was indeterminate :
since a 45' shear wave transducer was not used by GE on the lower side of H1.
- 2) Indication #1 between lugs 3 and 4 is ID surface connected and was not !
previously reported by GE in the RFO%109R1 report. However a review - (Reference 8) demonstrates that the same indication was located in RFO B109R1, but was interpreted as geometry. ; Reference 4 does not specifically evaluate the two indications described above. The indications are both below the H1 weld and therefore the depth of the H1 reinforcement fillet weld leg cannot be added to the remaining ligament for structural evaluation (i.e., the preservice section thickness must be considered to be 1.50 inches in lieu of 2.25 inches). ' L e
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. Carolina Power & Light Company ESR # 95-00765 l Brunswick Nuclear Plant Revision 1 ATTACHMENTE l Page 4 of 6 :
'l This analysis considers the cracks below weld H1 in the same terms as the !
i
,. reference'4 analysis, j l
k Averans FInw nanth The average flaw depth was determined by summing the individual crack depth measurements (for the flaws identified below H1) and dividing by the number of j measurements. UT data sheets (Reference 7) indicate that at least 710 millimeters (mm) of circumference was examined during RFO B110R1 and data was recorded
.f at least every five (5) mm. Therefore, approximately 142 readings (710/5) were )
4 taken. Of these, less than 25 showed flaws below H1 and the total summed flaw i depth for these readings was less than 147 mm. Accordingly, the average flaw ' depth for a postulated equivalent flaw extending around the entire shroud , i circumference below H1 is less than 0.05 inches (147mm/142 readings = 1.04 ! mm < 0.05 inches). ! I Flaw Grnwth Rate 1 The B110R1 inspection data was compared with B109R1 data to determine actual crack growth rates. Based on data comparkon, and as shown graphically in - Attachment D to ESR #95-00765, crack growth is substantially less than postulated in Reference 4. Indication #2 between lugs 26 and 27 was not previously reported by GE in the RFO B109R1 report, however a review performed by GE (Reference 6) demonstrates that the same indication was located in RFO B109R1 and showed an approximate depth of 0.45". The RFO B110R1 UT data indicates a maximum i depth of 0.472". Thus, the observed crack growth during Fuel Cycle #9 was A i i l
, , . . . , - ._ . . - _ _ _ _ _ . , _ _ .I
t t . Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 1 ATTACHMENTE Page 5 of 6 4 0.022". This is significantly less than the maximum equipment uncertainty of the I UT equipment used in RFO B109R1 and RFO B110R1, as shown in the following chart: ! REFUEL OUTAGE SURFACE CONNECTION EQUIPMENT UNCERTAINTY B109R1 ID or OD 0.1
- 9 B110R1 ID O.028" B110R1 OD 0.011" H1 crack growth can be conservatively bounded by equipment uncertainties and is considered to be 0.1" per fuel cycle for the purposes of this evaluation.
Enfatv Far tnr Datarminatinn Analysis of the surface connected flaws below H1 for the next two fuel cycles is summarized in Table 4 below: a "Waq@tm q i g g : 7+ gs j g_gd 3=gggggggsmap' egg g 3
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Upset 1.50 0.05 0.10 0.25 1.25 1.03 kai 48.8 2.25 l Faulted 1.50 0.05 0.10 0.25 1.25 1.86 ksi 26.9 l 1.125
j-g . Carolina Power & Light Company ESR # 95-00765 Brunswick Nuclear Plant Revision 2 A TTACHMENTE Page 6 of 6 i SLIMMARY The remaining net section thickness area projected for the section immediately below the H1 weld after the next two fuel cycles will be greater than that acceptably evaluated in Reference 4 (Compare Table 4 to Table 2). Therefore, the Reference 4 l evaluation fully bounds the identified condition below the H1 weld. The postulated ! crack lengths at the end of Cycle 11 (RFO B112R1) will not reduce the structural '! design margins below allowable values. Therefore, the evaluated condition does not impose any restrictions to BNP-1 operation during the next two operating cycles. REFERENCES ., 1. GE Report GENE-523-123-0993, Rev. 2, " Evaluation and Screening Criteria for the Brunswick 1 Shroud Indications," November 1993.
- 2. CP&L Calculation No.1-B21-0049, Revision 0
- 3. EER 93-0536, Evaluation of Unit 1 Core Shroud Indications and Operability Assessment of Unit 1 and 2, Revision 1
- 4. GE Report GENE-523-144-1093, " Analysis of Unit 1 Welds H1, H2, & H3...,"
] Revision 1, November 1993
- 5. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI,1980 Edition through Winter 1981 Addenda
- 6. Memorandum, A. R. Jaschke (GE) to John Langdon (CP&L), " Review of Unit 1 1993 Shroud Weld H1 UT Data," dated May 5,1995
- 7. RFO B110R1 IVVI and UT reports, including OPT-90.1 data sheets and video a I cassenes.
- 8. Memorandum, E. Black (CP&L) to W. B. Wilton (CP&L), " Review of H1 Weld UT Indication at 30 Degree Azimuth", dated April 17,1995 i
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