ML20072M846
| ML20072M846 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 06/04/1994 |
| From: | Gore P CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20072M829 | List: |
| References | |
| 94-0077, 94-0077-R00, 94-77, 94-77-R, NUDOCS 9409020207 | |
| Download: ML20072M846 (48) | |
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B N P RECtPIENT ID m(WyntasUn Form 2 ENGINEERING EVALUATION REPORT Page 1 EER Number Rev.
C-list System File Number Reference 94-0077 0
Class A 1005 Document (s)
G0250C Title Evaluation of Unit 2 Core Shroud Indications and Operability Assessment CLASSIFICATION (refer to Section 6.2 for additionalinstructions)
Seismic Related?
Yes / No EQ Affected?
Yes / No MOV per ENP 56?
Yes /g Follow NED seismic Design Guides; Follow PLP-02; Al-71 EQ Review; Consult NED Mech. or Component NED management approval ENP-34.1, Form 3 if required Ena. to address GL 89-10 Use As Is?
Yes / No Parts Upgrade?
Yes / No Short Term Structure' Yes / No Follow PMC 15.6 Integrity?
Expiration Date N/A Permanent Repair?
Yes / NJ Temporary Repair?
Yes/A Temporary Modification?
Yes /3 PLP-08 if ISl; Expiration Date N'A Follow PLP-22 Form 6 for Dwa/ Doc chances PLP-08 if ISI: Notify Temp Cond Coord Notify Temp Cond Coerd 50.59 Required?
6 / No FSAR Affected?
Yes /3 Operability Assessment?
M /No Cornplete Al 109 Safety Review; PNSC Two tech reviews if Class A; Complete within 0104 time frame;
& RCI-2.1 if unreviewed safety RCI-4.1 form; Corp. NSRG approval T/S management approval Approvals Print Name Signature Dgis Engineer Steve Bertz
[
4 t/
1st Tech Review 6f required)
Steve Bertz 84/
2nd Tech Review lif reauired)
Roger Steckel k
Manager - (title) Technent Supprirt Chip Pardee M
M M%
Others fspecifvl Principal Enar. - NED Paul Caiarella
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63 9/
System Enoineer Phil Gore lJb b gte 6 f MN
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ADDITIONAL DISTRIBUTION (beyond normal EER distribution per NRCS)
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List of Effective Pages/ Attachments
'JUN 0 719%
Pages 1-48 Revision 0
-- eu rer51 O ENP-12 36 UN,5 P' "
Rev.34 9409020207 940824 PDR ADOCK 05000324 P
Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision O i
ENGINEERING EVALUA TION REPORT Page 2 of 48 SECTION 1.0 TABLE OF CONTENTS Page #
TRAVELER.
1 (Section 1)
TABLE OF CONTENTS..
.............2 (Section 2)
EXECUTIVE
SUMMARY
4 i
(Section 3)
SHROUD DESIGN DISCUSSION.
......5 (Section 3.1)
Shroud Design.....
...5 (Section 3.2)
Miscellaneous Parts and Accessories
...7 (Table 3.1)
Shroud Weld Details.......
............8 (Section 4)
SHROUD FABRICATION AND INSTALLATION 11 l
(Section 5)
CAUSAL FACTORS..
12 (Section 6)
INSPECTION RESULTS..
13 (Table 6.1)
Detailed Inspection Results 15 (Section 7)
ANALYSIS AND RESULTS 17 (Table 7.1)
Analysis and Results Summary 18 (Table 7.2)
Screening Evaluation of H4 90 Circumferential Sector With the Most Indications...
22 i
(Table 7.3)
Screening Evaluation of H5 90 Circumferential Sector With the Most Indications.
23 (Section 8)
EVALUATION AND
SUMMARY
25 (Section 9)
REFERENCES 27 EER ACTION ITEM NOTIFICATION FORM.
28 SAFETY REVIEW (PER Al-109) 29 l
EQ IMP.ACT EVALUATION (PER ENP-34.1, FORM 3) 40 i
4 Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant.
Revision O ENGINEERING EVALUA TION REPORT Page 3 of 48 FIGURES (1)
Reactor Vessel Cross-Section Showing Reactor Internals...
41 (2)
Reactor Shroud Three-Dimensional View
..............42 (3)
Roll-Out View of inside Shroud Surface...............
43 (4)
Roll-Out View of Outside Shroud Surface.............
44 (5)
Brunswick Shroud Plan View......................45 (6)
Shroud Cross-Section Showing Welds.....
..........46 (7)
Separator Support Ring and Attachments............... 47 (8)
Core Spray Sparger Bracket........................ 48 0
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s Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 4 of 48 SECTION 2.0 EXECUTIVE
SUMMARY
in October,1990, RICSIL No. 054 reported cracking near the circumferential seam weld at the core beltline area of the shroud in a GE BWR/4 located outside the United States. Based on recommendations contained in this RICSIL, the BNP Unit 2 shroud was inspected during the 1991 refueling outage. No cracks were identified. The Unit 1 shroud 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. The tapes of the Unit 2 shroud IVVI were re-examined based on the early July Unit 1 findings. Three small indications were noted. Unit 2 tapes were again examined in late September, based on lessons learned on Unit 1. One additional small indication was noted.
Although this additional indication was bounded by the assumptions in the original evaluation, it was recognized that the quality of the 1991 tapes was insufficient to identify all of the types of cracks being confirmed on Unit 1. EER 93-0536 was issued to assess Unit 1 shroud structuralintegrity and to justify continued operation of Unit 2 until a detailed inspection could be performed during the Spring, 1994 RFO.
The Unit 2 inspections are complete and evaluated in this EER. This EER concludes that structural integrity of the core shroud will be maintained, with full FSAR safety margins, for at least the next 600 days of hot operation, and for welds H1, H4, i
and H5, for at least the next 1200 days of hot operation based on analysis of the inspection results. These durations allow operation at least through the next operating cycle, currently scheduled to end 2/2/96. However, the inspection plan for the next RFO will consider not only these inspection results, but will also consider continuing developments in the industry, to ensure utilization of the best i
information and technology to address the issue.
The inspection results confirm that the comparison presented in EER 93-0536 accurately reflected that Unit 2 was bounded by the Unit 1 analysis, and that continued operation was justified.
This EER is Quality Class A due to the quality classification of the shroud.
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s Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 5 of 48 SECTION 3.0 SHROUD DESIGN DISCUSSION The design of the reactor vessel internals was in accordance with applicable portions of Section lit of the ASME Boiler and Prc,5urs Vessel Code,1965 Edition through Summer,1967 Addenda (ref. 9 *). Although the shroud itself is not a Code component, the above Code wu used as the design basis for determining limits for stress intensities.
SECTION 3.1 Shroud Design The core shroud is a cylindrical assembly inside the reactcr vessel, which provides 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 burier external to the reactor vessel.
b)
Limit deflections and deformations of the reactor vessel internals 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 reactor vessel shroud support ring. The three regions are of different diameters: the top region is approximately 15'-9" diameter; the central region is approximately 14'-9"; and the lower region is tapered from 14'-9" to 14'-3" (see Figures 1 and 2, and ref. 9.2). Roll out maps of the core shroud depicting the locations of horizontal welds H1-H9, vertical welds V1-V11, plates P1-P11 and shroud hold down bolt lugs are shown in Figures 3,4, and 5. The weld and plate designations were assigned for inspection purposes.
The upper shroud consists of the separator support ring, the upper shroud cylindrical shell, and the top guido support ring. The separator support ring is constructed from 6 ring segments having a cross section of approximately 6" X 6", cut from rolled and annealed plate, welded together, then machined to final dimensions. The ring materialis Type 304 stainless steel from three different heats, with a carbon content of 0.047 - 0.061 wt% (The ring j
material in Unit 1 was from a single heat, different from the above 3, with 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.
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Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 6 of 48 The shell is formed from (2) 1 %" thick semicircular plates, welded together using a Double-U prep. The carbon contents range from 0.049-0.060 wt%.
These plates are from the same heats of material as the Unit 1 plates. The Top Guide Support Ring, with a cross section of 7%" X 3", is constructed and welded (H2) to the upper shroud shell in a manner similar to the separator support ring. The ring material has a carbon content of 0.064 wt%. The 6 ring segments were fabricated from a single heat of material, which was also used for 3 of the Unit 1 ring segments. These welds are oriented such that the axial residual stresses pull across the short transverse orientation (end grain) of the ring material.
The central region of the shroud consists of the mid-shroud barrel, the core support ring, and adjoining welds. The barrelis 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 0.046-0.061 wt%.
The mid-shroud barrel is welded to the upper shroud assembly at H3, which consist of a Single-J prep weld 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 at material. These are the same 2 heats used for the Unit 1 ring segments.
The lower region of the shroud consists of the lower shell course, shroud support ring, jet pump diffuser ring, and associated welds. The lower shell course is forrned from (3) 1 %" thick plates welded together using Double-U prep welds to form a conical section. Carbon contents range from 0.046-0.058 wt%. It is joined to the core support ring at weld H6b, which is sirnilar to H6a. The shroud support ring, which transfers the shroud weight and other loads to the reactor vessel, is 2" thick 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 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 l
reinforcement.
i Refer to Table 3.1 and Figure 6 for shroud welds and materials detail.
Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 7 of 48 SECTION 3.2 Miscellaneous Parts and Accessories A number of miscellaneous parts and accessories are associated with the core shroud. The following listing describes some of these components and their function and location.
SECTION 3.2.1 Separator Support Ring and Attachments (Figure 7)
The Shroud Head Bolt Lugs are located around the circumference of the Separator Support Ring. There are 36 pairs of lugs that align with matching lugs on the Shroud Head. The Head Bolts secure the shroud head to the shroud by applying a clamping force between the two sets of bolt lugs.
Therefore, the lugs function to provide the loading surfaces for the bolts.
SECTION 3.2.2 Shroud Access Manway Covers The Shroud Access Manway Covers provided access to the bottom head plenum during installation of the vessel internals. The cover is an Alloy 600 plate that fits into an opening in the Jet Pump Diffusser Ring at the 0 and 180o azimuths (directly below the recirculation suction nozzle). The access cover is welded in place using the shielded metal arc welding process (SM AW). Once the access cover is welded in place, the annulus between the shroud and the reactor vessel wall is secured for jet pump operation.
SECTION 3.2.3 Core Spray Spargers and Brackets The core spray spargers and brackets are located inside the core shroud between welds H1 and H2. The brackets are 304 stainless steel, welded to the spargers and the core shroud, and provide alignment and support for the spargers. BWR's have experienced cracking in the spargers and have been inspecting in accordance with IEB 80-13. A recent inspection at another utility found cracking at the brackets supporting the spargers. BNP expanded its IVVI plan to incorporate these brackets.
SECTION 3.2.4 Jet Pump Beam Riser Braces The jet pump riser braces are made of 304 stainless steel and welded to the reactor vessel on one end and the riser pipe on the other end. They are located at the upper part of the jet pump assembly. The braces provide lateral support to maintain jet pump alignment and structural integrity during operation. The jet pump beam riser braces were in the original vessel internals inspection plan as recommended by GE SIL 551.
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8 Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision O ENGINEERING EVALUA TION REPORT Page 8 of 48 TABLE 3.1 - SHROUD WELD DETAILS COMPONENT DESCRIPTION PIECE NUMBERS
UPPER SHROUD SEPARATOR SUPPORT RING Piece 5 304 SS SSR assembled from (SSR) 0.047 - 0.061 wt%
six plate segments, welded with 308L SS Double-U welds l
WELD H1 Double-J:
308L SS ID welded first, OD Fillet on ID back-chipped, then welded UPPER SHROUD Piece 1 304 SS Assembled from 2 SHELL COURSE 0.049 - 0.060 wt%
rolled plates, welded together by 308L SS Double-U welds V1 and V2 l
WELD H2 Double-J; 308L SS ID welded first, OD '
Fillet on ID back-chipped, then welded TOP GUIDE Piece 6 304 SS TGSR assembled SUPPORT RING 0.064 wt%
from six plate (TGSR) segments, welded I
with 308L SS Double-U welds WELD H3 Single J on ID:
308L SS ID welded first, OD Fillet on OD back-chipped, then welded i
Carolina Power & Light Company
. EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 9 of 48 TABLE 3.1 -~ SHROUD WELD DETAILS j
COMPONENT DESCRIPTION PIECE NUMBERS
MID-SHROUD BARREL MID-SHROUD Piece 2 304 SS Assembled from 2 I
TOP SHELL COURSE (Upper) 0.051 -0.058 wt%
rolled plates, welded together by 308L SS Double U welds V3 l
and V4 WELD H4 Double-J 308L SS One of last two welds made to assemble shroud.
MID SHROUD Piece 3 304 SS Assembled from 2 MIDDLE SHELL COURSE 0.046 - 0.061 wt%
rolled plates, welded together by 308L SS Double U welds V5 and V6 WELD H5 Double-J 308L SS One of last two welds made to assemble shroud.
MID-SHROUD Piece 2 304 SS Assembled from 2 LOWER SHELL COURSE (lower) 0.051 - 0.056 wt%
rolled plates, welded together by 308L SS Double-U welds V7 and V8 l
WELD H6a Double-J; 308L SS Fillet on ID CORE SUPPORT RING Piece 7 304 SS CSR assembled from (CSR) 0.063 - 0.067 wt%
six plate segments, welded with 308L SS Double-U welds WELD H6b Double Bevel; 308L SS Fillet on ID
Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision O ENGINEERING EVALUA TION REPORT Page 10 of 48 TABLE 3.1 - SHROUD WELD DETAILS COMPONENT DESCRIPTION PIECE NUMBERS
LOWER SHROUD LOWER SHROUD Piece 4 304 SS Assembled from 3 TAPERED SHELL COURSE 0.046 - 0.058 wt%
rolled plates, wolded together by Double-U welds V9, V10, V11 WELD H7 Single Bevel on Alloy 82 root Gas Tungsten Arc ID; Fillet Welded Welded (GTAW)
Backing Ring Ailoy 182 filler Root: Shielded Metal on OD Arc Welded (SMAW)
Fill SHROUD SUPPORT RING N/A Alloy 600 Plate thickness is (SSR) 2.0" 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 DIFFUSER RING 2.5" WELD H9 Double-J Alloy 82 root GTAW root.
(attaches Jet Pump with Fillets SMAW fill.
Diffuser Ring to Reactor Alloy 182 filler Vessel)
- Sun Shipbuilding & Dry Dock Fabrication piece reference numbers.
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i Carolina Power & Light Company EER # 94-0077 Brunswick Nuc! ear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 11 of_48 SECTION 4.0 SHROUD FABRICATION AND INSTALLATION The core shroud was designed by General Electric and fabricated by Sun Shipbuilding & Dry Dock Company from January 1970 to June 1971 for Unit 2 and January 1970 to November 1971 for Unit 1. The Unit 2 core shroud was installed in October 1973, and the Unit 1 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 (Ref. 9.3). No significant fabrication or installation details were discovered that would indicate any material conditions unique to either of the units. However, the weld material used to assemble the shrouds for each Unit was different.
The weld material was 308 SS for Unit 1 and 308L SS for Unit 2, for both the circumferential welds (H1, H2, H3, H4, H5, H6a, and H6b) and the vertical welds j
(V1 - V11) using automatic Submerged Arc Welding (SAW) and/or manual Shielded Metal Arc Welding (SMAW) procedures. This low carbon weld material has no appreciable effect on the resistance to IGSCC since the welding process is the same and the plate material is Type 304 in both Units, i
Welds H7, H8, and H0 are constructed of similar materials and processes on both Units. There are no significant differences.
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Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision O ENGINEERING EVALUA TION REPORT Page 12 of 48 i
SECTION 5.0 CAUSAL FACTORS The factors that affect IGSCC and their relation to the core shroud are detailed in EER 93-0356. Unit 2 was compared to Unit 1, to demonstrate that the histories of both units are similar. This comparison considered water chemistry, shroud I
materials and fabrication techniques, and critical hours of operation. Cracking histories of other components are also compared.
Both Units had variances in conductivity in the early years, followed by a trend of smaller variances and lower conductivity for recent years. Unit 1 has geometric j
and calculated mean conductivities of 0.333 pS/cm and 0.959 S/cm, respectively, i
Unit 2 has corresponding values of 0.328 S/cm and 0.915 S/cm.
The Hydrogen Water Chemistry (HWC) system for Unit 2 has more cumulative operation than Unit 1. HWC mitigates the environmental conditions that are favorable to crack formation and growth.
No significant differences were identified during the review of the materials, fabrication and installation techniques for the core shrouds for Unit 1 and Unit 2.
Cumulative hot operating hours on both units are within one operating cycle of j
each other, with Unit 2 having a slightly longer total operating time.
)
A comparison.was made of IGSCC experience for. recirculation system pipe and the shroud head bolts. Accounting for recirculation system pipe replacement and on-line months, Unit 1 and Unit 2 experienced similar cracking events in the recirculation system piping and shroud head bolts.
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t Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Phnt Revision 0 ENGINEERING EVALUA TION REPORT Page 13 of_48 SECTION 6.0 WNSPECTION RESULTS Inspections of various components within the reactor are routinely performed each refueling outage in ac:ordance with the requirements of ASME Code Section XI, and the plant in-Service Inspection (ISI) Program.
The inspection plan for Unit 2 was based on the experience and observations from Unit 1 and the current understanding of the causal factors contributing to the cracking being seen. Accordingly, the initial inspection plan provided extensive inspections for the regions where the most cracking had been observed (above the core plate (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 to be performed in accordance with GENE-523-123-0993, Revision 2, " Evaluation and Screening Criteria for the Brunswick.1 Shroud Indications," dated 11/93 (Reference 9.9). The design criteria and safety margins in this document are the same for both Units and, therefore, this document is applicable to Unit 2. Visual examinations involved cleaning the weld area 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 in order to ensure detection of tight cracks.
The inspection plan was revised to reflect the decision to install the modification at H2/H3, and to take advantage of weld specific analyses that had baen performed -
RAM-94-092/ SIR-94-029, " Addendum to the Brunswick Unit 1 Screening Criteria" dated 4/6/94 (Reference 9.1); and RAM-94-099/ SIR-94-031, " Minimum Required Unflawed Core Shroud Material at Brunswick, Units 1 and 2, dated 4/11/94 (Reference 9.11). A sampling inspection for weld H1 established four symmetric inspection windows picked to cover 4 of the 6 ring segments, and at least 2 of the 3 material heats used to fabricate the ring. 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.
The inspection plan was again revised during the course of the shroud inspections, to take advantage of the successful utilization of the automated ultrasonic scanner at another utility. The scanner offered the advantage of providing both length and depth characterization of detectable cracking, with only an OD inspection and without having to clean the welds. The automated scanner worked reasonably well on the H4 weld, but was unable to function properly on the H5 weld and could not be used. H5 was inspected visually.
A final change was made to the inspection plan due to reports of cracking at the lower core plate found at another utility. The sample plan for welds below the core plate was doubled to include areas at both O' and 180 Sample locations were chosen to cover at least 4 of the six ring segments comprising the lower core support ring, and at least one sample from each heat of material used to fabricate the ring.
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- Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 14 of 48 The results of the inspection found no indications in the areas sampled on H1, and in' the lug welds examined. H2 was inspected at the area of previously identified indications, and cracking was determined to extend across all of the area inspected.
H4 and H5 indicated moderate degrees of cracking.- Minor cracking was seen in the H6a and H6b areas, and no cracking seen on H7. Based on the observation at H6a -
H7, absence of any indications at the access hole covers, and simi'ar findings on Unit 1, H8 and H9 were not inspected. Table 6.1 provides a detailed account of the inspection findings. Reference 9.12 contains the specific Inservice Inspection results.
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Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision O ENGINEERING EVALUA TION REPORT Page 15 of 48 TABLE 6.1 - UNIT 2 DETAILED INSPECTION RESULTS WELD RESULTS H1 VISUAL INSPECTIONS: 75 -85, 105 -175*, 255*-265, 345 -355 OD areas were inspected on the top and bottom of the weld. Additional weld metal,39.5" top and 7" bottom toa of the weld, was inspected as part of the shroud hold down bolt lug weld inspection. A total of 17.7% of the top and 12.3% of the bottom of the weld was inspected. No indications were identified.
UT INSPECTIONS:
None performed.
H2 VISUAL INSPECTIONS: 40*-50 OD area was inspected to confirm four indications identified from a previous RFO tape. A circumferential crack was identified in the TGSR that ran continuously through the inspection zone. The other indication, in the plate, was determined to be in-line pitting.
UT INSPECTIONS:
None performed.
H3 VISUAL INSPECTIONS: None performed.
UT INSPECTIONS:
None performed.
H4 VISUAL INSPECTIONS: 350"-10 OD area was inspected and no indications were identified.
UT INSPECTIONS:
Approximately 78% was inspected from the OD using an automated UT device. 23 circumferential indications were ider tified,9 in the top toe and 14 in the bottom toe. Lengths of indications ranged from 0.3" to 13.6" The depth ranged from 0.10" to 0.86" with an average of 0.53" Cracking was in the heat affected zone of the weld.
H5 VISUAL INSPECTIONS: 97.8% of ID was inspected with 7 circumferential and 3 axial cracks identified. The longest circumferential cracks ranged from
').25 to 12" The axial cracks were less than 1.25" long.
30.6% of the OD was inspected with 1 circumferential and 2 axial crrcks identified. The circumferential crack was 11" long.
The axials were less than 0.5" All circumferential cracking was in the heat affected zone of the weld.
U1 INSPECTIONS:
None performed.
Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision O ENGINEERING EVALUA TION REPORT Page 16 of 48 TABLE G.1 - UNIT 2 DETAILED INSPECTION RESULTS WELD RESULTS H6a VISUAL INSPECTIONS: 350 -10 and 170 -190 areas (11 %) were inspected from the OD. 4 axial cracks, 2" and less in length, were identified in the and CSR. One circumferential crack 1.5" long was identified in the plate above H6a, in the heat affected zone.
None performed.
H7 VISUAL INSPECTIONS: 350 -10 and 170 -190" areas (11%) were inspected from the OD. No indications were identified.
UT INSPECTIONS:
None performed.
H8 VISUAL INSPECTIONS: None perform 0d.
UT INSPECTIONS:
None performed.
H9 VISUAL INSPECTIONS: None performed.
UT INSPECTIONS:
None performed.
V1 - V11 VISUAL INSPECTIONS: None performed.
UT INSPECTIONS:
None performed.
PLATES VISUAL INSPECTIONS: None performed.
UT INSPECTIONS:
None performed.
ATTACH-VISUAL INSPECTIONS: Additional inspections performed include:
MENTS
- Top Guide Upper and Lower Beams. No reportable indications.
and
- Shroud Head Bolt Lugs (8 pairs on the Shroud). No reportable indications.
COMPON-
- Manway Access Hole Covers (Oo and 180 azimuths). No ENTS reportable indications.
- Jet Pump Beam Riser Braces. No reportable indications.
- Core Spray bracket welds. No reportable indications.
- C arc Spray Sparger has a crack emanating from a flow nozzle tack weld.
UT INSPECTIONS:
UT inspections were performed on the Manway Access Hole Covers and no indications were noted.
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Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 17 of 48 SECTION 7.0 ANALYSIS AND RESULTS The cracks were analyzed for structural significance by initially screening them in accordance with Reference 9.9. The screening process conservatively assumes that the cracks are through-wall, and provides guidance that is more conservative than ASME Code Section XI criteria by combining the lengths of cracks that are relatively close together (effective crack length). It then provides a bounding crack length for initial screening. Cumulative effective crack lengths which are smaller than the bounding crack length are not a structural concern and are screened from a specific evaluation. Effective crack lengths that are larger than the bounding crack length must be specifically analyzed.
For the purposes of this analysis, the Unit 2 core was assumed to be critical 7/2/94 and run continuously until the next refueling outage 2/2/96. This is approximately 580 days. The analysis used 600 days in assessing crack growth to bound this
~
period.
BNP also developed a weld-specific structural analysis to determine more precisely the allowable crack size (Reference 9.10 and 9.11). The allowable flaw in a vertical weld is 106". Since this allowable length is less than the width of any of the shroud
[
assembly plates, vertical welds were not included in the inspection plan.
i Table 7.1 is a detailed account of the analysis results for each weld joint. Table 7.2 and 7.3 provide a summary of H4 and H5 cracking for comparison to the Screening Criteria.
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Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision O ENGINEERING EVALUA TION REPORT Page 18 of 48 TABLE 7.1 - UNIT 2 ANALYSIS AND RESULTS
SUMMARY
WELD DISPOSITION ANALYSIS AND RESULTS H1 SPECIFIC Four 10 symmetric locations were inspected with no cracks ANALYSIS identified. This amounts to 16.5" of uncracked metal at each
]
PERFORMED location. The minimum required ligament at each location is 7.25" to justify operation for one fuel cycle. Each 600 day operating cycle has an assumed crack growth of 1.44". At the end of one 600 day operating cycle, any cracking in the j
uninspected areas would not grow to sufficient length to violate design margins. This weld requires no further inspection and is qualified for a minimum of one 600 day operating cycle.
H2 REPAIR 40" - 50* VT indicated that the weld had significant cracking in the area inspected.
H3 REPAIR N/A c-H4 ACCEPTABLE The summed effective lengths of circumferential cracks in the BY limiting 90 sector after 600 days of operation is 52.2". UT SCREENING depth measurements ranged from 0.10" and 0.86" Table 7.2 l
CRITERIA summarizes the most limiting 90* sweep of circumferential cracks on H4.
H5 ACCEPTABLE The summed effective lengths of circumferential cracks in the BY limiting 90 sector after 600 days of operation is 61.4". The i
SCREENING axial cracks were less than 1.25" Table 7.3 summarizes the CRITERIA most limiting 90 sweep of c{cumferential cracks on H5.
H6a ACCEPTABLE Inspections on the OD identified 4 short axial indications in the BY CSR. No indication exceeded 2" in length. One circumferential j
and SCREENING crack 1.5" long was identified in the plate. This is consistent CRITERIA with the cracking identified on Unit 1, where similar water H6b chemistry, material, and fabrication processes exist. The cracking identified in the 40 of inspection is assumed to be representative of the remaining uninspected areas, for analysis purposes. The total projected cracking is 13.5 (1.5"
- 360 /40"). The separation between cracks is projected to be at least 10, so the maximum effective crack length in any one 90 sector after one 600 day operating cycle is expected to be
]
26.5" (9 ' [1.5 " +.72" +.72") = 26.5"). This is less than the 74.5" allowable and passes the screening criteria.
)
i H7 ACCEPTABLE No indications were identified in either the backing ring fillet or BY H7 weld. This inspection is supported by the detailed inspection SCREENING of the manway access hole covers. No indications were CRITER!A identified by VT or UT. (See the addendum at the end of this table for more detailed discussion).
Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision O ENGINEERING EVALUA TION REPORT Page 13 of 48 TABLE 7.1 - UNIT 2 ANALYSIS AND RESULTS
SUMMARY
WELD DISPOSITION ANALYSIS AND RESULTS H8 N/A Not inspected. This is an inconel weld. This weld is similar to Unit 1 H8. in material and fabrication, where no cracking was identified. The Manway Access Hole Covers are in the shroud support plate and are also inconel welds but in addition have a crevice. This design is more susceptible to'lGSCC by crevice corrosion end is expected to show evidence of cracking before H8. The VT and UT inspection of these covers found no indications. Therefore, the H8 weld was not expected to be cracked and was not inspected.
H9 N/A Not inspected. This is an inconel weld. This weld is similar to Unit 1 H9, in material and fabrication, where no cracking was identified. The Manway Access Hole Covers are in the shroud support plate and are also inconel welds but in addition have a crevice. This design is more susceptible to IGSCC by crevice corrosion and is expected to show evidence of cracking before j
H9. The VT and UT inspection of these covers found no indications. Thercfore, the H9 weld was not expected to be cracked and was not inspected.
V1 - V11 N/A Not inspected. The allowable axial flaw size to maintain j
structural integrity is greater than the width of the widest plate.
Therefore, the vertical flaws are bounded by analysis, and are not inspected.
PLATES N/A Not inspected. Only one crack was found in the Unit 1 plate.
The IGSCC rate used in the analysis is 5.0 E -5 in/hr or j
0.72"/600 days of operation. Since plate cracks are expected to be the result of local cold work, a plate crack will not have a j
predetermined direction, if a crack initiated at plant start-up and grew for life of the plant, it would be less than 2' O.72"(600 days / cycle)
- 40 cycles or 57.6". This is less than the allowable length for horizontal and axial cracks in the screening criteria.
ATTACH-As Noted No cracking was observed in the MENTS
- manway access hole covers
- separator hold down bolt lugs and
- jet pump beam riser braces
- top guide upper and lower beams COMPON-
- core spray sparger brackets ENTS Cracking was identified in a core spray sparger flow nozzle. EER 94-0137 was issued which evaluated the crack as acceptable with a follow-up inspection at the next RFO.
Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Ruvision O ENGINEERING EVALUA TION REPORT Page 20 of 48 ADDENDUM TO TABLE 7.1: DISCUSSION OF WELD H7 The juncture of the core shroud (1 %" thickness) and the shroud support ring (2" thickness) is a field installation weld known as H7. This weld is different from the other shroud welds for several reasons as follows:
Welding process is different (GTAW and SMAW versus SAW).
Dissimilar metal weld (Type 304 stainless steel /A!!oy 600). Alloy 600 backing ring applied to outer surface of joint and lef t in-place after welding. Technique different from conventional backing ring approach in that the ends are tapered towards the shroud, a fillet weld is used to seal the upper end of the backing ring (SMAW process with Alloy 182 filler), and the bottom of the backing ring is fused to the Alloy 600 shroud support ring using the GTAW process (Alloy 82 filler added). Proper fusion using this method eliminates the potential for a crevice on either edge of the backing ring. The weld cavity was filled from the 10 using the SMAW process (Alloy 182 coated electrodes)
Liquid penetrant examination was performed on the root weld and the completed weld and on the final surf ace of the backing ring fillet weld.
j No grinding could have been made to the outer surface after welding because nf restricted
- access, i
Similarities to other welds included substantial fit-up stresses and probable grinding of the ID weld crown af ter completion of welding.
This weld is difficult to inspect from the inside of the shroud due to restricted access through core structures, it is dif ficult to inspect from the outer surface because inspection is restricted by jet pump diffusers except for the limited area below the suction lines. The presence of the backing ring is a limiting factor on the outer side as well.
The concern for this weld is that the Alloy 182 filler material is susceptible to IGSCC, especially if the location is creviced.
The H7 we!d is believed to be without cracking for the following reasons:
j No indications were seen in the IVVI inspection performed on the outer surface at the 0 and 180 azimuths (11% of shroud periphery). Extra care was taken with cleaning and with positioning of the camera to maximize quality of the inspection in addition, visual evidence of uniform backing ring melting and tie-in to the shroud support ring indicates either that no crevice exists or at least the crevice characteristics are minimal.
The coolant has a low oxidizing power on both sides of the shroud at this location. This means that the environment is much less aggressive than it is higher on the shroud.
The shroud manway access covers were inspected by IVVI and ultrasonic inspection during the current outage. The shro d manway access covers had no reportable indications. The shroud manway ac ass cover is more constrained than the H7 weld and
4 e r j
Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 21 of_48 I
has higher residual stress. Additionally, the welds for the covers are ground, creviced and use the same w elding process as the H7 weld. The shroud manway access cover welds are therefore a conservative comparison for the H7 weld.
There were no reportable indications for the H8 and H9 welds on Unit 1.
?
Nickel based materials are highly resistant to attack in chloride environments. Since chlorides were the major source of the high coolant conductivity for the first five years of.
operation, the nickel based materials would be very resistant to crack initiation during that period. As discussed in the root cause section, this was believed to be a factor in the extent of cracking seen higher on the shroud. Therefore, the H7, H8, H9 and access cover welds would have been more resistant to degradation from the high conductivity condition in the early years of operation.
The root of the H7 weld was made using the GTAW process. Alloy 82 filler material was used for the first three passes of the weld. The weld was completed using the SMAW process using Alloy 182 filler metal. This means that the Alloy 182 materialis not exposed to the coolant on the shroud OD except at the fillet on the upper side of the backing ring. This weld is low stress because of the low degree of constraint. Alloy 82 materialis known to be highly resistant to IGSCC because of a higher chromium content than Alloy 182. The Alloy 182 material on the shroud ID is not creviced.
CONCLUSION:
The H7 weld is not believed to be cracked based on an valuation of the factors above.
l f
1
Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision O ENGINEERING EVALUA TION REPORT Page 22 of 48 l
TABLE 7.2 H4 EVALUATION IND.
AZIMUTH AZIMUTH LENGTH EFFECTIVE REMAINING THRU-WAL ORIENTATION START SToP in LENGTH LIGAMENT L
DIMENStoN 1
33 37.5 7.0 15.3 1.09 0.41 circ upper 2
35 42.3 11.2
.74 0.76 cire lower 3
93.3 102 13.6 15.1 1.29 0.21 circ upper 4
108.1 113.3 8.0 9.5 1.09 0.41 circ upper 5
131.7 133.7 3.2 1.00 0.50 circ upper 6
135.3 142.9 11.8 25.5 0.92 0.58 circ upper 7
143.9 147.2 4.4 0.64 0.86 cire upper 8
199.4 200.3 1.4 2.9 1.37 0.13 cire upper 9
208.5 211.9 5.4 6.9 1.09 0.41 cire lower 10 220.3 223.0 4.1 5.6 0.85 0.65 circ lower 11 231.0 231.5 0.7 2.2 1.28 0.22 circ upper 12 241.5 244.0 3.8 10.5 0.79 0.71 circ lower 13 245.2 247.1 3.0 0.71 0.79 circ lower 14 249.9 252.1 3.3 10.1 0.76 0.74 cire lower 15 252.3 255.5 4.8 0.71 0.79 circ lower 16 261.2 262.3 1.7 6.2 1.04 0.46 circ lower 17 263.5 264.2 1.0 1.40 0.10 circ upper 18 275.6 275.8 0.3 1.8 1.38 0.12 cire lower 19 299.2 300.9 2.6 4.1 0.93 0.57 circ lower 20 308.4 310.2 2.5 4.0 1.00 0.50 circ lower 21 316.1 323.5 11.5 13.0 0.65 0.85 circ lower 22 329.9 332.4 l 3.8 5.3 0.79 0.71 cire lower 23 340.6 343.9 l 4.5 6.0 0.77 0.73 circ lower
)
OD 350 010 none
-VT 113.60 143.9
\\
4 Indications 12 through 22 combine for the longest total effective crack in a 90 sector, 52.2", at the end of one 600 day operating cycle. This is less than the allowable flaw size of 74.5" in a 90 sector i
in the screening criteria. therefore, the weld is qualified for at least one 600 day operating cycle.
9 Carolina Power & Light Company EER # 94 0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 23 of 48 TABLE 7.3 H5 EVALUATION CELL AZIMUTH AZIMUTH LENGTH EFFECTIVE CRACK CRACK ORIENTATION START STOP in LENGTH START STOP ummmmmmmmummuummmmmmu ammmmmmmmmmmmmmmmmmummummmuum-30-51 005 013 1.5 012 012 axial 34-51 013 025 none 38-47 025 035 none 42-47 035 045 0.5 041 041 axial 46-43 045 055 none 46-39 055 065 none 50-35 065 075 none 50-31 075 085 0.25 1.7 084 084 cire 50-27 085 095 none 50-23 095 105 none 50-19 105 115 none 46-15 115 125 none 46 11 125 135 none 42-07 137 147 0.5 138 138 axial 38-07 145 155 0.25 148 148 axial 34-03 155 167 1.0 2.5 163 163 cire 34-03 155 167 1.0 164 164 axial 34-03
- 155, 167 1.0 156 156 axial 30-03 167 175 0.25 174 174 axial 30 03 167 175 9.3 168 175 circ 26-03 175 185 l
5.0 24.8 175 178 cire 26-03 175 185 l
4.5 179 183 circ 22-03 185 193 2.0 187 187.5 cire (below) 22-03 185 193 3.0 10.8 190 192 circ 22-03 185 193 l
3.5 186 188 circ 18-03 193 203 2.5 4.0 197 199.5 circ 14-07 210 215 3.0 4.5 210 212 cire 10-07 215 223 none 06-11 227 235 0.25 3.8 2 30.E.
230.5 circ 06-11 227 235 1.0 231.5 232 cire 06-15 235 240 l
none 02-19 248 260 12.0 13.5 249 257 cire 02-23 257 267 0.25 262.5 262.5 axial 02-23 257 267 1.5 3.0 260 261 circ 02-27 265 275 none 02-31 275 283 none 02-35 283 295 none 06-39 295 305 1.0 297 297 axial
Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision O ENGINEERING EVALUA TION REPORT Page 24 of 48 TABLE 7.3 H5 EVALUATION CELL AZIMUTH AZIMUTH LENGTH EFFECTIVE CRA?K CRACK ORIENTATION START STOP in LENGTH STAAT STOP mmm m
06-43 303 315 6.0 7.5 308 312 cire 06-43 303 315 0.5 2.0 305 305 circ (below) 10-47 315 325 7.5 9.0 318 322.5 cire 14-47 325 330 1.2 325 325 axial 18-51 335 345 none 22-51 345 355 none 26-51 355 005 none OD 010 020 none OD 040 050 none OD 070 080 none OD 100 110 none OD 130 140 none OD 160 170 none OD 220 230 0.25 227 227 axial OD 220 230 0.38 226 226 axial OD 280 290 none OD 340 350 11.0 12.5 341 348 circ OD 350 010 none ummmmmmmm-l l 81.88 l 99.6 Indications between 168' and 257 combine for the longest effective crack,61.4", at the end of 600 days of operation. This is less than the allowable flaw size of 74.5" in a 90 sector in the screening criteria, therefore, the weld is qualified for at least one 600 day operating cycle.
J Carolina Power & Light Company EER # 94-0077
\\
Brunswick Nuclear Plant Revision O ENGINEERING EVALUA TION REPORT Page 25 of 48 SECTION 8.0 EVALUATION AND
SUMMARY
The Unit 2 circumferential cracking was " tight" and followed the HAZ of the horizontal welds. There was evidence of heavy machining and grinding at the welds. The UT reflectors at H4 were indicativa of IGSCC flaws. The cracking had characteristics similar to that found in Unit 1 where metallography confirmed the cracking to be IGSCC. Therefore, the root cause of cracking in the Unit 2 core shroud is believed to be the same root cause as in Unit 1, i.e., IGSCC.
Based on similarity of fabrication and operation with Unit 1, Unit 2 was expected to be susceptible to similar cracking in the core shroud welds. An inspection plan was developed around the Unit 1 Screening Criteria (ref. 9.9). From experience in Unit 1, it was recognized that all areas of the welds would not be accessible for inspection. Each weld was evaluated for the most appropriate inspection technique and the appropriate sample size to qualify the core shroud for at least one 600 day operating cycle.
H1 is accessible for VT from the OD but the ID configuration does not permit camera and lighting angles to readily detect IGSCC cracking. A sampling strategy similar to that of ASME Section XI was used, in that approximately 10%
of the weld was selected for VT. There were four sample sites symetrically located about the circumference.
If the cracking was as extensive as found in Unit 1, the inspection scope would expand to the balance of the weld. No indications were identified. To provide additional assurance of structural integritry, a weld-specific analysis was performed using the inspection results.
The analysis concluded that, if all uninspected areas are assumed cracked H1 would maintain full structural design margins for at least one 600 day operating cycle.
H2 was inspected at the 40 -50 OD area to confirm two indications identified from a previous RFO tape. A circumferential crack was identified in the TGSR that ran continuously through the inspection zone. The other indication, in the plate, was determined to be in-line pitting. H3 was not inspected, since modification of the shroud was planned (PM 94-007).
H4 was inspected by an initial VT from the OD at 350 -10 in conjunction with a UT from the OD at all accessible areas. The UT provided data which demonstrated that the weld was qualified by the Screening Criteria for at least one 600 day operating cycle. (The cracks were conservatively as.sumed through-wall for the Screening Criteria analysis).
H5 was inspected for approximately 100% by VT from the ID. VT on the OD was performed on accessible areas which were not cracked on the ID. The UT device was not able to track behind the jet pumps due to tight clearences and therefore was not used. The VT provided data which demonstrated that the weld was qualified by the Screening Criteria for at least one 600 day operating cycle.
]
H6a and H6b were inspected at two locations and compared to the inspection results from Unit 1. The cracking was primarily axial with one short circumferential crack. This is similar to the cracking experienced in Unit 1, which was fabricated from the same heats of materials as the Unit 2 ring. The cracking
- )
Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 26 of 48 was assumed to be representative of the uninspected areas and was qualified by the Screening Criteria for at least one 600 day operating cycle.
H7 was inspected at two locations and compared to the inspection results from Unit 1. No indications were found at H7 in Unit 1. H7 is not as susceptible to IGSCC as the Access Hole Covers (AHC). A VT.and UT inspection found no indication at the AHC welds, which further confirms that the H7 weld is not cracked..The weld was qualified by the Screening Criteria for at least one 600 day operating cycle.
Welds H8 and H9 were not inspected based on the lack of indications in Unit 1 and the results from the AHC inspection. The weld was qualified by the Screening Criteria for at least one 600 day operating cycle.
The vertical welds were not inspected. The allowable axial flaw size to maintain structural integrity is greater than the widest plate..Therefore, the vertical flaws are bounded by analysis.
The surface area of the plates was not inspected. If a crack were to have initiated at plant start-up and grow for life of the plant, it would be less than 2*
0.72"(600 days / cycle)
- 40 cycles or 57.6". This is less than the allowable length for horizontal and axial cracks in the screening criteria. Therefore, the plate surfaces are bounded by analysis, i
1 1
I i
Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 27 of 48 SECTION
9.0 REFERENCES
9.1 Brunswick Updated FSAR, Table 3.9.5-6 9.2 FP-50096, Sheet 1 of 2, " Assembly and Finish Machining Shroud Core Structure," Revision 2.
9.3 Technical Memorandum, " Comparison of Brunswick Units 1 & 2 Core Shroud Fabrication and Installation," TM-B-1005-003, dated October 21,1993.
9.4 EER 93-0536, Evaluation of Unit 1 Core Shroud Indications and Operability Assessment of Unit 1 and 2.
9.5 GE Report NEDC 32300-P, " Brunswick Unit 1 Shroud Sample Metallurgical Evaluations," dated October 1993.
9.6 Fluence Reference for Shroud OD and ID: Flux at Shroud Wall Versus Azimuth, Tables, October 16,1993.
9.7 Westinghouse Report WCAP-10903, " Reactor Cavity Neutron Measurement Program For Carolina Power And Light Company Brunswick Unit 2," December 1986.
9.8 EPRI Report NP-944, " Studies on AISI Type-304 Stainless Steel Piping Weldments for Use in BWR Applications," December 1978.
9.9 GE Report GE-NE-523-123-0993, Rev. 2, " Evaluation and Screening Criteria for the Brunswick 1 Shroud Indications," November 1993.
9.10 RAM-94-092/ SIR-94-029, " Addendum to the Brunswick Unit 1 Screening Criteria", dated 4/6/94.
9.11 RAM-94-099/ SIR-94-031, " Minimum Required Unflawed Core Shroud Material at Brunswick, Units 1 and 2", dated 4/11/94.
9.12 OPT-90.5, in-Vessel Visual Examination, Rev 11, dated 3/25/94.
l 9.13 System Description SD-01, " Nuclear Boiler," Revision 26, dated Nov.1,1993.
F Carolina Power & Light Company EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 28 of 48 i
Form 3 ENGINEERING ACTION ITEM Page 1 EAl Number 94-0077-1 Priority 3b Due Date 2/10/96 i
Assigned Manager Paul Caf arella Concurrence Name/Date Paul Cafarella - 6/3/94 Action Required in January,1996, review the B212R1 schedule against the analysis period of this EER (600 days with f
the core critical) to assure that any changes are bounded. Issue a revision if necessary.
Originator / Group (print)
Ext.
Date Supervisor Signature /Date Tracking Entry Steve Bertz 3182 06-03-94 (initials)
Resolution O Close; provide documentation O Transfer to
- obtain concurrence O Extend to
- provide basis (PGM approval to extend temp conditions
)
Responsible Individual Signature /Date Responsible Manager Signaturn/Date ROUTE TO TECHNICAL SUPPORT ENGINEERING DATA COORDINATOR Tracking Updated (initialsi 0-ENP-12 37 Rev. 34
_ _. _ ~, -
ATTACHMENT 2 (Cont'd)
EER 94-G077 Rev. O Page 29 REVISION 3 10CFR50.59 PROGRAM MANUAL ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page _,1_ cf 11 SAFETY REVIEW COVER SHEET DOCUMENT NO. EER 94 0077 REV.NO.
O DESCRIPTION OF TITLE: Evaluation of U2 Core Shroud Indications and O. A.
- 1. Assigned Responsibilities:
Safety Analysis Preparer:
Steven L. Bertz Lead 1st Safety Reviewer:
Steven L. Bertz 2nd Safety Reviewer:
Roaer Steckel
- 2. Safety Analysis Preparer: Complete PART l. SAFET NALYSIS leMF Safety Analysis Preparers
/
Date
- 3. Lead 1st Safety Reviewer: Complete Part II, item Classification.
- 4. Lead 1st Safety Reviewer: ll1 may be completed. If either question 1 or 2 is "yes," then Part IV is not required.
- 5. Lead 1st Safety Reviewer: Determine which DISCIPLINES are required for review of this item (including own) and mark the appropriate blocks below.
DISCIPLINES Reauired:
IPrint Name)
Sianature/Date (Steo 7)
[ ] Nuclear Plant Operations
[ ] Nuclear Engineering M
[X) Mechanical Steven Bertz N #fM4[hY/
I ] Electrical
[ ] Instrumentation & Control X] M allurgy Steve Williams w LW M7 M[
[ ] Chemistry /Radioc' maistry l ] Health Physics
[ ] Administrative Controls j
- 6. A QUALIFIED SAFETY REVIEWER will be assigned for each DISCIPLINE marked in step 5 and his/her name printed in the space provided. Each person shall perform a SAFETY REVIEW and provide input into the Safety Review Package.
- 7. The Lead 1st Safety Reviewer will assure that a Part 111 or Part IV is completed (see step 4 above) and a Part VI if required (see 9.d of Part II) j Each person listed in step 5 shall sign and date next to his/her name in 1
step 5, indicating completion of a SAFETY REVIEW.
8,2nd Safety Reviewer: Perfor SAFETY VIEW in accordance with Section 8.0 2nd Safety Reviewer
'M Date 6 4-DISCIPLINE:
( / Mechanical Yes,M i
- 9. PNSC review required? If "yes" attach Part V and mark reason []
below:
[ ] Potential UNREVIEWED SAFETY QUESTION
[ ] Question 9 of Part IV answered "Yes"
[ ] Other (specify):
0 Al-109 Rev. 002 Page 72 of 86
ATTACHMENT 2 (Cont'd)
EER 94-0077 Rev. O Page 30 REVISION 3 10CFR50.59 PROGRAM M ANUAL ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page _2_ of 11 PART 1: SAFETY ANALYSIS (See instructions in Section 8.4.1)
(Attach additional sheets as necessary)
DOCUMENT NO. EER 94-0077
.REV. NO.
O DESCRIPTION OF CHANGE:
Based on RICSIL No. 054, the BNP Unit 2 shroud was inspected during the 1991 refueling outage. No cracks were identified. The Unit 1 shroud was inspected in July,1993, and a near 360a circumferential crack was confirmed on the inside diameter of the Top Guide Support Ring, in the heat affected zone of the weld.
Additional in-Vessel Visual inspections (IVVI) were conducted, and confirmed additional circumferential and axial indications elsewhere in the shroud on both the inside and outside diameter. The tapes of the Unit 2 shroud IVVI were re-examined based on the July,1993 Unit 1 findings. Three smallindications were noted.
The indications were assumed to be cracks (although not confirmed) and were conservatively evaluated in Engineering Evaluation Report 93-0477. Unit 2 tapes were again re-examined in late September, based on lessons learned on Unit 1. One additional smallindication was noted. Although this additionalindication was bounded by the assumptions in the original evaluation,it was recognized that the quality of the 1991 tapes was insufficient to identify all of the types of cracks being confirmed on Unit 1.
Unit 2 has completed a inspection of the core shroud welds using visual and ultrasonic techniques. The observed cracking was similar to Unit 1 with the exception that no cracking was observed at H1 or the shroud head lug welds. H2 was inspected at one location to confirm indications called from previous outage tape.
No further inspections were performed on these welds since they will be repaired prior to start-up. The comparison presented in EER 93-0536 accurately reflected that Unit 2 was bounded by Unit 1.
The purpose of this EER is to evaluate the significance of cracking observed in the Unit 2 shroud with respect to operation of the unit for one 600 day operating cycle.
ANALYSIS:
Tne reactor internals perform the following safety related design basis functions as specified in the UFSAR:
1.
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.
2.
Limit deflections and deformation to assure that the control rods and the core standby cooling systems can perform their saf ety functions during abnormal operational transients and accidents.
3.
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.
Intergranular stress corrosion cracking (IGSCC) of the type and form experienced with recirculation piping and related systems in Boiling Water Reactors (BWRs) is the cause of cracking. Crack extension is possibly assisted by neutron fluence and " oxide wedging" at certain locations. Susceptible material conditions, high 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 the shroud, the location and degree of cracking varies across the shroud.
i 4
4 ATTACHMENT 2 (Cont'd)
EER 94-0077 Rev. O Page 31 l
REVISION 3 10CFR50.59 PROGRAM MANUAL ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page 3. of 11 PART I: SAFETY ANALYSIS (CONT'D.)
ANALYSIS (Cont'd.):
The core shroud must maintain a floodable volume above the two-thirds core height elevation. The cracks are
~
caused by intergranular stress corrosion cracking, and inherently are tight. Any through wall cracks would result in negligible leakage into the downcomer region and be contained by the reactor pressure vessel. The Emergency Core Cooling systems provide suf ficient make-up and cooling capacity to ensure that the fuel will remain covered.
i EER 93-0536 was issued to assess Unit 1 shroud structuralintegrity and to justify continued operation of Unit 2 until a detailed inspection could be performed at the next RFO. The Unit 2 inspection is complete and this EER provides results of the ana!ysis of the cracking on the Unit 2 shroud. Welds H4, H5, H6a, H6b, and H7 meet the original sc;eening criteria and will remain within the criteria for at least one 600 day operating cycle, i
The screening criteria assumes that the cracks are through-wall, and provides guidance that is more conservative than ASME Code Section XI criteria. Effective crack lengths which are smaller than the screening criteria are not a concern and require no further evaluation. Effective crack lengths that are larger than the screening criteria must be specifically analyzed. A location specific analysir was performed for weld H1 demonstrating structuralintegrity for at least two 600 day operating cycles (maximum recommended by SIL 572). H2 and H3 did not receive a fullinspection because a permanent mechanical repair of these Top Guide Support Ring welds will be made prior to Unit 2 startup (this avoids continued inspection and evaluations at this area). H8 and H9 were not inspected based on finding no indications in the access hole cover we!ds, and similar findings on Unit 1.
Structural integrity of the core shroud will be maintained, with full FSAR safety margins, for a minimurn of one 600 day operating cycle based on analysis of the inspections performed. Permanent mechanical.epair of the H2 and H3 weld areas at the Top Guide Support Ring will be made prior to Unit 2 startup. Thia avoids continued inspection and evaluations at this area.
l 4
4 i
J v
e--
=-
ATTACHMENT 2 (Cont'd)
EER 94-0077 Rev. 0 Page _3JL REVISION 3 10CFR50.59 PROGRAM MANUAL i
ATTACHMENT A 1
CP&L SAFETY REVIEW PACKAGE PageSof11 PART 1: SAFETY ANALYSIS (CONT'D.)
ANALYSIS (Cont'd.):
REFERENCES:
The Brunswick Updated FSAR was reviewed for information on the design functions of the shroud. Numbers in parentheses () refer to UFSAR paragraph number.
(3.2.1)
Structures and equipment are classified as Seismic Class 1 if they are essential for safe shutdown or if their failure could result in the release of radiation with dose consequences potentially exceeding the guidelines of 10CFR100.
(3.2.1.2)
The core shroud is classified as Seismic Class 1.
(3.9.2.5.1) The reactor core structural components are designed so that deformations produced by accident loadings do not prevent insertion of control rods.
(3.9.5.1)
The core shroud is a part of the reactor vesselinternals. The core shroud up to the level of the jet pump nozzles is a part of the floodable inner volume of the reactor vessel.
(3.9.5.2.1) The following load combinations and safety factors were used:
l 1.
The OBE plus upset pressure difference load combination should be evaluated with a safety factor of 2.25.
2.
The DBE plus normal operating pressure difference load combination should be evaluated with a safety factor of 1.50.
3.
The load combination of DBE plus LOCA plus normal loads should be evaluated with a safety factor of 1.125.
4.
The load combination of LOCA plus normalloads should be evaluated with a safety factor of 1.50.
(3.9.5.3)
The design of th9 reactor vessel internals wac in accordance with applicable portions of the ASME B & PV Code Section ill 1965 edition through Summer 1967 Addenda.
NOTE: There are no applicable portions of Section ll1 for the core shroud.
Where applicable Codes and Standards did not exist, the reactor vesselinternals were designed to the criteria in Section 3.9.5.2 and to the limits in Tables 3.9.5-1 through 3.9.5-4.
i (3.7.1.1.2) The DBE ground horizontal acceleration is 0.16g. The vertical DBE ground acceleration is equal to two thirds of the horizontal acceleration. OBE ground acceleration is one-half of the DBE accelerations.
REFERENCES:
FSAR Chapters 1.2.2.5.11, 3.2.1.2, 3.9.2.5.1, 3.9.5.1, 5.3.1, 5.3.3.1.2.3, 5.4.1, i
ATTACHMENT 2 (Cont'd)
EER 94-0077 Rev. O Page 33 j
REVISION 3 10CFR50.59 PROGRAM MANUAL i
ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page f of 11 1
PART 1: SAFETY ANALYSIS (CONT'D.)
ANALYSIS (Cont'd.):
REFERENCES:
FSAR Chapters 1.2.2.5.11, 3.2.1.2, 3.9.2.5.1, 3.9.5.1, 5.3.1, 5.3.3.1.2.3, 5.4.1, 7.3.3.1.3.5, 7.7.1.1.2.2, 7.3.3, 9.3.4.2, Ch.15: Tech Spec. 3/4.3.3 and associated i
- basis, j
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ATTACHMENT 2 (Cont'd)
EER 94-0077 Rev. O Page 34 REVISION 3 10CFR50.59 PROGRAM MANUAL ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page _6__, of 11 PART 11: ITEM CLASSIFICATION DOCUMENT NO. EER 94-0077 REV.NO.
0 Yes No
- 1. Does this item represent:
- a. A change to the facility as described in the SAFETY
[]
IX]
ANALYSIS REPORT 7
- b. A change to the procedures as described in the
[]
[X]
SAFETY ANALYSIS REPORT?
- c. A test or experiment not described in the SAFETY
[]
IX]
ANALYSIS REPORT?
- 2. Does this item involve a change to the individual plant
[]
[X]
Operating License or to its Technical Specifications?
- 3. Does this item require a revision to the FSAR?
[]
[XI
- 4. Does this item involve a change to the Offsite
[]
[X]
Dose Calculation Manual?
- 5. Does this item constitute a change to the Process Control
[]
[XI Program?
- 6. Does this item involve a major change to a Radwaste Treatment
[]
[X]
System?
- 7. Does this item involve a change to the
[]
[X]
Technical Specification Equipment List?
- 8. Does this item impact the NPDES Permit (all 3 sites) or
[]
[XI constitute an "unreviewed environmental question" (SHNPP Environmental Plan Section 3.1) or a "significant environmental impact" (BSEP)?
- 9. Does this item involve a change to a previously accepted:
- a. Quality Assurance Program
[]
[X)
- b. Security Plan (including Training,
[]
[XI Qualification, and Contingency Plans)?
- c. Emergency Plan?
[]
[X]
- d. Independent Spent Fuel Storage installation license?
I]
[X]
(if yes, refer to Section 8.4.2, " Question 9," for special considerations. Complete Part VI in 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 1-9 above, identify specific reference sections used for any "Yes" answer.
See Safety Evaluation references.
O Al-109 Rev. 002 Page 74 of 86
ATTACHMENT 2 (Cont'd)
EER 94-0077 Rev. O Page 35 REVISION ?
10CFR50.59 PROGRAM MANUAL ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page 1 of 11 PART lil: UNREVIEWED SAFETY QUESTION DETERMINATION SCREEN DOCUMENT NO. EER 94-0077 REV.NO.
0
~
YES NO
- 1. Is this change fully addressed by another completed
[]
[XI UNREVIEWED SAFETY QUESTION determination? (See Section 7.2.1, 7.2.2.5, and 7.9.1.1)
REFERENCE DOCUMENT:
REV.NO.
YES NO
- 2. For procedures, is the change a non-intent change which only
[]
[X]
(check all that apply): (See Section 7.2.2.3)
[ ] Correct typographical errors which do not alter the meaning or intent of the procedure; or, i 1 Add or revise steps for clarification (provided they are consistent with the original purpose or applicability of the procedure); or,
[ ] Change the title of an organizational position; or,
[ ] Change names, addresses, or telephone numbers of persons; or,
[ ] Change the designation of an item of equipment where the equipment is the same as the original equipment or is an authorized replacement; or,
[ ] Change a specified tool or instrument to an equivalent substitute; or,
[ ] Change the format of a procedure without altering the meaning, intent, or content; or
[ ] 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 111is "Yes," then PART IV need not be completed.
O Al-109 Rev. 002 Page 75 of 86
o' ATTACHMENT 2 (Cont'd)
EER 94-0077 Rev. 0 Page 36 REVISION 3 10CFR50.59 PROGRAM MANUAL ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page J_ cf 11 PART IV: UNREVIEWED SAFETY QUESTION DETERMINATION DOCUMENT NO. EER 94-0077 REV.NO.
0 1
Using the SAFETY ANALYSIS developed for the change, test or experiment, as well as other required references (LICENSING BASIS DOCUMENTATION, Design Drawings, Design Basis Documents, codes, etc.),
the preparer of the SAFETY EVALUATION must directly answer each of the following seven questions and make a determination of whether an UNREVIEWED SAFETY QUESTION exists.
A WRITTEN BASIS IS REQUIRED FOR EACH ANSWER Yes No
- 1. May the proposed activity increase the
[] [X]
probability of occurrence of an accident evaluated previously in the SAFETY ANALYSIS REPORT?
See attached.
- 2. May the proposed activity increase the consequences of an
[] [X]
accident evaluated previously in the SAFETY ANALYSIS REPORT?
See attached.
- 3. May the proposed activity increase the probability of
[] [X]
occurrence of a malfunction of equipment important to safety evaluated previously in the SAFETY ANALYSIS REPORT 7 See attached.
- 4. May the proposed activity increase the consequence
[] [X]
of a malfunction of equipment important to safety evaluated previously in the SAFETY ANALYSIS REPORT?
See attached.
- 5. May the proposed activity create the possibility
[] [X) of an accident of a different type than any evaluated previously in the SAFETY ANALYSIS REPORT?
See attached.
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0 Al-109 Rev. 002 Page 76 of 86
ATTACHMENT 2 (Cont'd)
EER 94 0077 Rev. O Page 37 REVISION 3 10CFR50.59 PROGRAM MANUAL ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page 9 of 11 PART IV (Continued)
DOCUMENT NO. EER 94-0077 REV.NO.
0 Yes No
- 6. May the proposed activity create the possibility of a
[] [X]
malfunction of equipment important to safety of a different type than any evaluated previously in the SAFETY ANALYSIS REPORT?
See attached.
- 7. Does the proposed activity reduce the margin of safety as
[] [X]
defined in the basis of any Technical Specification?
See attached.
- 8. Based on the answers to questions 1 - 7, does this item
[] [X]
result in an UNREVIEWED SAFETY QUESTION? If the answer to any of the questions 1-7 is "Yes", then the item is considered to constitute an UNREVIEWED SAFETY QUESTION.
- 9. Is PNSC review required for any of the following reasons?
[] [X]
If, in answering questions 1 or 3 "No", it was determined that the probability increase was small relative to the uncertainties; or, in answering question 2 or 4 "No", it was determined that the doses increased, but that the dose was stillless than the NRC ACCEPTANCE LIMIT; or in answering question 7 "No", a parameter would be closer to the NRC ACCEPTANCE LIMIT, but the end result was still within the NRC ACCEPTANCE LIMIT; then PNSC review is required.
REFERENCES:
See Safety Evaluation references.
This Unreviewed Safety Question Determination is for the following DISCIPLINE (s): (Additional Part IV forms j
may be included as appropriate.)
[ ] Nuclear Plant Operations
[ ] Structural
[ ] Nuclear Engineering
[X] Meta!!urgy
[X] Mechanical
[ ] Chemistry / Radiochemistry
[ ] Electrical
[ ] Health Physics
[ ] instrumentation & Control
[ ] Administrative Controls O AI-109 Rev. 002 Page 77 of 86
EER 94-0077 Rev. O Page 38 PART IV: UNREVIEWED SAFETY QUESTION DETERMINATION (Cont'd.)Page /4 'of 11 DOCUMENT NO. EER 94-0077 REV.NO.
0 1.
May the proposed activity increase the probability of occurrence of an accident evaluated previously in the SAFETY ANALYSIS REPORT?
The core shroud cracking patterns evaluated in this EER will not increase the probability of occurrence of an accident as defined in the UFSAR. The EER demonstrates the structuralintegrity of the core shroud and affirms that the core internal geometric alignment will be maintained. Therefore, structural integrity margin is maintained throughout the cycle of operation.
2.
May the proposed activity increase the consequences of an accidant evalurted previously in the SAFETY ANALYSIS REPORT 7 The core shroud cracking patterns evaluated in this EER will not increase the consequences of an accident previously evaluated in the UFSAR. The core shroud must maintain a floodable volume above the two-thirds core height elevation. The cracks are caused by intergranular stress corrosion cracking, and inherently are tight. Any through wall cracks would result in negligible leakage. The ECC systems provide sufficient make-up and cooling capacity to ensure that the fuel will remain covered. The EER also demonstrates the structuralintegrity of the core shroud which insures that the core geometry will be maintained. Maintenance of core alignment and floodable volume assure that design basis will be maintained.
3.
May the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAFETY ANALYSIS REPORT 7 The core shroud structuralintegrity is not compromised by the presence of the cracks evaluated in j
the EER. The depth of the cracks projected at the end of the operating cycle will be less than the allowable crack size. Therefore, structural integrity margin is maintained throuahout the cycle of operation.
4.
May the proposed activity increase the consequence of a malfunction of equipment important to safety evaluated previously in the SAFETY ANALYSIS REPORT 7 The core shroud cracking pattern described in the EER will not increase the consequence of a previously evaluated malfunction of equipment importent to safety. The core shroud functions to l
maintain a floodablo volume above two-thiro core height, and to maintain core alignment to insure control rod insertion. This function is assured since structuralintegrity is demonstrated. In addition, the mitigating functions of the ECC system will not be measureably affected by any reasonably assumed leakago through the cracks.
l 5.
May the proposed activity create the possibility of an accident of a different type than any evaluated previously in the SAFETY ANALYSIS REPORT 7
EER 94-0077 Rev. O Page 39 PART IV: UNREVIEWED SAFETY QUESTION DETERMINATION (Cont'd.)Page j,L' of 11 DOCUMENT NO. EER 94-0077 REV.NO.
O The cracking pattern described in the EER will not create the possibility of an accident different than any evaluated previously in the UFSAR. The structuralintegrity will be maintained, assuring alignment of the core internals. The ability of the control rods to insert will not be impaired. The cracks are IGSCC which inherently provide a narrow torturous path for leakage. Any through wall cracks would result in negligible leakage. The ECC systems provide sufficient make-up and cooling capacity to ensure that the fuel will remain covered.
6.
May the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAFETY ANALYSIS REPORT 7 The cracking pattern described in the EER will not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the UFSAR. Maintaining structural integrity ensures that the core shroud performs its functior.at requirements and thus no equipment important to safety will be adversely influenced and no new failure modes will be introduced.
7.
Does the proposed activity reduce the margin of safety as defined in the basis of any Technical Specification?
The cracking pattern described in the EER does not reduce the safety margin as defined in the Technical Specification Bases. Structural integrity will be maintained which assures the ability to insert the control rods and maintain a floodable volume.
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ENGINEERING EVALUI. TION REPORT EER 94-0077 Rev. O ENVIRONMENTAL QUALIFICATION IMPACT FORM (EER-EQIF).
Page 40 Will the evaluation, on either a temporary or permanent basis:
1 1.
Justify the deletion of equipment / common components from the BSEP EQ program?
[] Yes
[X] ~ No j
2.
Justify the addition of (already existing) equipment / common components to the BSEP EQ program?
[] Yes
[X] No 3.
Authorize the repair of EQ equipment / common components with other than cualified like-in-kind equipment / components parts?
[] Yes
[X] No 4.
Affect the existing installation or interface (of EQ equipment / common component applications) as may be designated in EDBS and/or in the qualification data package (including changing the type of interface / installation)?
[] Yes
[X] No.
5.
Justify the (quality class) upgrade of equipment / common components p2 component parts which could be utilized in EQ applications?
[] Yes
[X] No 6.
(Re) Define qualification parameters (e.g.,
normal or LOCA/HELB environmental conditions, post-accident operating time requirements, essential passive / active post-accident operating requirements, qualified life assumptions /results,'etc.) for specific EQ equipment?
[] Yes
[X] No 7.
Provide an EQ-related justification for continued operation (as required per PLP-02, Section 4.4.3.3 pl 4.4.4)?
[] Yes
[X] No 8.
Provide the resolution of a qualification problem (as required per PLP-02, Section 4.4.4)?
[] Yes
[X] No Notes:
1.
If all,no, then no further EQ consideration is required.
Mark the EER Traveler accordingly as required by ENP-12 and include this completed EER-EQIF within the EER package. An EQ Technical Review is not required.
2.
If any yes, an EQ impact assessment (per Section 5.3) must be j
performed during the evaluation process.
Mark the EER Traveler accordingly and include this completed EER-EQIF within the EER package. An EQ technical review is required.
1 BSEP/Vol. XX/ENP-34.1 20 Rev. 4 i
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UNIT 0 Rev. 10/91 RMP-007 Ret: Life Page 1 of 1 ILIIGIBLE RECORD ACCEPTANCE FORM originator requests acceptance of this document and accepts responsibility for the illegible condition of this data.
?
Document Identity EcC @d- 077 kN ()
m
. < cia (6 i
EQTI-The identity of the illegible record or illegible page(s) within the record shall be provided by the originator of this form by identifying and inserting this form preceding the illegible data.
In cases where the entire record is considered illegible, this form precedes the record.
Please complete Part A og,B.
)COr.X)uuuuuvot x x x x x x x x x x x_x xx xxx x-x x x x x x x x x xxxx4Ay 11 x x x xx x x x x x x x x x x x x x x x x AA)UUG.JUUUQUygtx x x x_Axxy x x xx;yyyytx_x x x x x x x xx x xx_m x x x x x x_xxx x x x x_xAAAxx.x x x x'r PART A The attached record is suitable for microfilming because:
It is non-Q or non-vital records or the' data which is relevant to the identification of the item is legible and/or the data can be provided from other sources.
gf
/
Signed:.8/
IV
Title:
f>) &
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/ S /4 #' L 7/94 o
Date XAXXAAA.h~)UUQUgyytxxx_x x_wx x x x x x x x x x x x x x xx x xxx x xx x x u,Juyt_xx_x xmx_x x_rr XXAx4XXXXAAAA)Luxxxym x x_x u;uggggggggyutx_x x x x_x Agggutxxx x x xx x x uxxxAxy uxxAAAAA PART B The attached record is the most legible copy available and may be retained in I
the RFR.
/
Originating Supervisor Date
{
Reviewed:
l
/
Supervisor - Nuclear Records Date Management 0-RMP-007 Rev. 10 Page 20 of 23 l
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Carolina Power & Light EER # 94-0077 Brunswick Nuclear Plant Revision O ENGINEERING EVALUA TION REPORT Page 41 of 48 un i
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Carolina Power & Light EER # 94-0077 Brunswick Nuclear Plant Revision O ENGINEERING EVALUA TION REPORT Page 42 of 48 o
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, o Carolina Power & Light EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 45 of 48 i
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.c Carolina Power & Light EER # 94 0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Page 47 of 48
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Carolina Power & Light EER # 94-0077 Brunswick Nuclear Plant Revision 0 ENGINEERING EVALUA TION REPORT Paqe 48 of 48 LOCATION OF AN INDICATION APPPOXIPAATELY M" IfJ LENGTH IN SFARGEP. ORIGINATIf1G FPOM NOZZLE WELD 2W CORE SPRAY 'A' LOOP UPPER SPARGER f
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