ML20203M014
| ML20203M014 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 02/19/1998 |
| From: | Gore P, Murray W CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20203L987 | List: |
| References | |
| ESR-9700706, ESR-9700706-R01, ESR-9700706-R1, NUDOCS 9803060355 | |
| Download: ML20203M014 (23) | |
Text
6
(
Pcgs 1 of _
Form 1 ENGINEERING SERVICE REQUEST ESR # 9700706 Rev 8 1
WR/JO f Other Documents (CR, OEF, etc.)
Primary System 8 Primary System Name Plant / Unit BNP g Multiple Systems 005 B21,B11 NUCLEAR BOILER tlNC.RX VESSEL &
Attected Title Originator / Phone Urvt 1 Core Shroud inspection Results YEMMA, LAWRENCE P
/850-3144 Plant Customers (Pnnt Name, Sgn, Date)
Engineering / Plant Programs (Pnnt Name. Sign, Date) g Regulatory Affairs hl1u,MUAAAf //A M Reviews (Pnnt Name, Sgn, Date)
DesgnVenicaten phjk bgMQphD fiq[96 t MDIHW D Other Revwws Requiredi O Records Atth>wd Engineering Disciplines (Pnnt Name, Sgn Date)
FEB 191998 lillCLDA 00:Uls[IllCOM e
Response Type DC-EE ESR N/A Quetty Class Team A Safety-Related Due Date 02 20 98 APPROVALS is a 10CFR 60.69 Safety Rev6ew required per tplant specinc procedure)?
[ NAS Before Approval / implementation O Safety Evaluaton revsed.
(Print Narne, Sign. Date)
SC eA oval / implementation
[ NRC Before implernentation 1st OSR: [. N/P)Afd,/
///P!98 m,
Date:
0 riant Genersi Manaaer WO;ahg)-
()h A
Date: Jfld)h g 2nd OSR:
Responsible Engineer LAWRENCE P YEMMA e W
Responsible Manager ; (?rt't Name, Sqn, Date)
%h b
1L/IT[1g i
Plant General Manager (Pnnt Name, Sqn, Date)
Procedure Form EGR-NGGCM5-1-3 OCM01b2a 03/31/97 9903060355 990225 PDR ADOCK 05000325 P
. o
c Pegs 2 et Form 1 ENGINEERING SERVICE REQUEST LSR #.
How #
Tatie 9700706 1
Unit 1 Core Shroud inspection Results Request:
Demonstrate, using the results of the previous Unit 1 invessel core shroud inspections, that core shroud structural integrity can be maintained for a minimum of one additional operating cycle with sufficient margins.
f
Response
Response summary:
Unit 1 Core Shroud welds H4, H5, H6a, H6b, and H7 are qualified for at least two additional operating cycles based on this evaluation (see attached).
Weld H1 is qualified for at least one additional operating cycle based on this evaluation (see attached).
Based on these conclusions, no inspection of the Unit I core shroud welds needs to be performed during the B112R1 outage.
Revision 1 of this ESR was generated to incorporate editorial changes and clarifications as noted.
Procedure: Form EGR-NGOC-0005-13 DCM03 03,78/96
Ccrolina Power & Light Company ESR 9700706 Brunswick Nuclear Plant - Unit 1 Revision 1 Engincoring Evaluation Page 3 List of Effective Pager, Page Rev Page Rev Page Rev 1
1 2
1 3
1 4
0 5
1 6
1 7
0 8
0 9
0 10 0
11 0
12 0
13 0
ATTACHMENTS 14 0
A 0
15 0
B 0
16 0
C 0
17 0
D 0
18 0
E O
3
i Carolina Power & Light Company ESR 9700706 Brunswick Nuclear Plant - Unit 1 Revision 0 Engineering Evaluation Page 4 Table of Contents Section Pace CoverSheets.......................................................................................................
1 Li s t o f E f f e c t i v e P a g e s..........................................................................................
3 TableofContenis................................................................................................
4 1.0 Purpose.........................................................................................................
5 2.0 Conclusions...................................................................................................
5 3.0 Background...................................................................................................
5 4.0 F u n c ti o n al R e q u i r e m e n t s.................................................................................
6 5.0 References....................................................................................................
7
- 6. 0 D e s i g n B a si sl I n p u t s.......................................................................................
9 7.0 Evaluation...................................................................................................
10 8.0 Fluence.......................................................................................................
12 9.0 S a f ety Fa cto rs ( End of Evaluatio n)..................................................................
14 10.0 R ei n s p e c tio n R e q uir e m e n t s..........................................................................
15 1 1. 0 S a f e t y A n al y s i s........................................................................................
16 Attachments Description No. of Pages A
H1 Fillet Weld Measurement 2
B DLL Benchmark C 'iculation 9
C DLL Analysis Output 24 D
B112R1 Shroud Inspection Plan 6
E Sketch of Weld H1 Upper inspection Data 1
1 1
I
Ccrolina Power & Light Company ESR 9700706 Brunswick NucIcar Plant - Unit 1 Revision 1 Engineering Evaluation Page 5 1.0 PURPOSE This ESR Evaluation is required as part of Carolina Power and Light Company's commitmr.nt to USNRC Generic Letter 94 03 "Intergranular Stress Corrosion Cracking of Coro Shrouds in Boiling Water Reactors." As such, this evaluation provides the following:
A review of the B111R1 inspection data incorporating new design inputs and revised assumptions. Specifically, since the last analysis, the shroud horizontal weld stresses have been revised per ref. 5.21 and are incorporated into the analyses of this evaluation. Additionally, credit for the reinforcing fillet weld on the inside diameter of the shroud is taken in this evaluation. This fillet was measured during the last outage (B111R1), but tvas not considered in the previous analysis (ref. 5.18).
This evaluation provides a basis for the continued use'of1he Unit 1 core shroud for at least one additional operating cycle, in the as-analyzed condition, without any horizontal weld inspections this outage or any operational changes or restrictions.
This is accomplished by a review of the core shroud structural capacity to the criteria established by the BWRVIP (references 5.6, 5.7) and approved by the Nuclear Regulatory Commission (reference 5.11).
A basis for the determination of future shroud inspections.
2.0 CONCLUSION
S t
The Unit 1 shroud can safely operate in the present condition for a minimum of l one additional fuel cycle with full UFSAR safety margins, without any operational l changes or restrictions.
3.0 BACKGROUND
in October 1990, RICSIL 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 R!CSIL 054, the Unit 1 shroud was inspected in July 1993, and a near 360 degree circumferential crack was confirmed on the inside diameter of the top guide support ring, at the H2 and H3 welds. Welds H2 and H3 were permanently repaired by the installation of a series of clamps about the outside circumference of the shroud. Engineering evaluation EER 93-0536 (reference 5.4) was issued to assess Unit 1 shroud structuralintegrity.
The Unit 1 shroud was re-inspected during the B110R1 refueling outage in the spring of 1995. Engineering evaluation ESR 95-00765 documents the shroud inspection results.
This ESR concludes that structuralintegrity of the core shroud will be maintained with full
Corolina Power & Light Company ESR 9700706
(
Brunswick Nuclear Plant - Unit 1 Revision 1 Engineering Evaluation Page 6 UFSAR safety margins for at least the next two fuel cycles and that crack growth during cycle 9 was substantially less than postulated by previous analysis.
The Unit 1 shroud was re inspected during the B111R1 refueling outage in the fall of 1996. Engineering evaluation ESR 96 00160 documents the shroud inspection results.
This ESR concluded that structuralintegrity of the core shroud will be maintained with full UFSAR safety margins for at least two fuel cycles and that crack growth during cycle 9 was substantially less than postulated by previous analysis. The upper side of the H1 wold fell under the plant specific analysis category and was qualified for one cycle using the best information available at the time.
Based on the previous BNP shroud inspections for Units 1 and 2, BWR industry inspection experience, BWRVIP and NRC recommendations, and improved inspectico tooling, the B111R1 shroud inspection plan was submitted to the NRC on January 23,1998.
The B112R1 shroud inspection plan is included in this report as Attachment D. The inspection plan calls for no core shroud weld inspections for this outage (8112R1).
Justification for this plan is contained in this evaluation. As can be seen from the previous evaluation (ESR 96-00160), all welds, except H1 upper, were qualified for multiple cycles. The upper side of H1 is qualified for an additional cycle based on additional wold data (fillet weld) that was obtained but not used during the B110R1 outage. See ref. 5.6 Section 5 for details on use of fillet welds.
4.0 FUNCTIONAL REQUIREMENTS Details concerning the shroud design, fabrication and functional requirements can be found in reference 5.4. This ESR documents the following safety design basis functional requirements for the core shroud:
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, 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 (a) and (b) above are satisfied so that the safe shutdown of the plant and removal of decay heat are not impaired.
Carolina Power & Light Company ESR 9700706 Brunswick Nuclear Plant - Unit 1 Revision 0 Engincering Evaluation Page 7
5.0 REFERENCES
5,1 U.S. Nuclear Regulatory Commission Generic Letter, GL 94-03 "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors" dated July 25,1994 5.2 U.S. Nuclear Regulatory Commission Letter to Mr. R. A. Anderson of Carolina Power and Light Company dated January 3,1995 " Generic Letter 94 03, Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors, Brunsveick Steam Electric Plant, Units 1 and 2 (TAC NOS. M90084 and M900B5)" including Enclosure 1 " Safety Evaluation for Unit 1" and Enclosure 2
" Safety Evaluation for Unit 2" 5.3 Engineering Evaluation Report, EER 94-0077, " Evaluation of Unit 2 Core Shroud Indications and Operability Assessment." Revision o dated 06/04/94 5.4 Engineering Evaluation Report, EER 93-0536, " Evaluation of Unit 1 Core Shroud Indications and Operability Assessment of Unit 1 and 2" Revision 2 dated 03/13/94 5.5 Engineering Service Request, ESR 95-00765, " Unit 1 Core Shroud Reinspection Evaluation" Revision 2 dated 06/20/95 5.6 BWR Vessel Internals Project, BWRVIP-01, "BWR Core Shroc<f /nspection and Flaw Evaluation Guidelines" Revision 2 dated October 1996.
5.7 BWR Vessel Internal Project, BWRVIP-20, "BWR Core Shroud Distributed Ligament Length Computer Program" Version 2.1 dated December 1996 as modified 12-30-97.
5.8 BWRVIP inspection Committee document " Reactor Pressure Vessel and /nternals Examination Guidelines (BWRVIP-03)" TR-105696, dated October 1995 5.9 Structural Integrity Associates document RAM-94-092iSIR-94-029, " Addendum to the Brunswick Unit 1 Screening Criteria", dated April 6,1994.
5.10 CP&L Brunswick Nuclear Plant - Plant Operating Manual - Volume X - Period
-Test - OPT-90.1, "In-Vessel Visual Examinations", Rev 17, dated 10/03/96 5.11 U.S. Nuclear-Regulatory Commission Letter to Mr. J.T.- Beckham, Chairman BWRVIP, dated June 16,1995 " Evaluation of BWR Core Shroud /nspection and Evaluation Guidelines, GENE 523-113-0894, Revision 1, dated March 1995, and BWRVIP Core Shroud NDE Uncertainty & Procedure Standard, dated November 22,1954 "
4
Carolino Power & Light Comp:ny ESR 9700706 Brunswick Nuclear Plant - Unit 1 Revision 0 Engineering Evaluation Page 8 5.12 GE Document GENE-813 00715-1 Revision 0 " Brunswick Nuclear Plant - Unit 1
- Shroud Examination Plan" Dated: September 1996 5.13 Supplement 1 Report to OPT 90.1 - GE BNP B112R1 Inspection Summey dated October 1996.
5.14 Engineering Service Request, ESR 96-00154, " Unit 2 Core Shroud Reinspection Evaluation" Revision 0 dated 03/08/96 5.15 Drawing FP-50096 Sheet 1 of 2 " Core Structure Assy & Final Machine -
Shroud & Shroud Top" Revision 2 5.16 BWR Vessel Internals Project, BWRVIP-07 " Guidelines for Reinspection of BWR Core Shrouds," dated February 1996.
5.17 Engineering Evaluation Report, ESR 96-00154, " Evaluation of Unit 2 Core Shroud Indications and Operability Assessment," Revision o dated 03/08/96 (B212R1) 5.18 Engineering Evaluation Report, ESR 96-00160, " Evaluation of Unit 1 Core Shroud Indications and Operability Assessment," Revision 0 dated 10/31/96 (B111R1) 5.19 GE Nuclear Energy Document GENE-523-A028 0495, " Brunswick Units 1 and 2, and Hatch Units 1 and 2 Shroud Safety Assessment," Rev. O dated Dec 20, 1996.
5.20 CP&L Calculation 1821-1008. " Evaluation of Core Shroud Seismic Loads," Rev.
O dated Dec 10,1993.
5.21 CP&L Calculation OB2's-1012, " Calculation of Core Shroud Horizontal Weld Stresses," Rev. O dated Sept. 15,1997.
5.22 CP&L Ca\\cu\\ation 1 B21-0049,
" Application of GENE-523-123-0993,
" Evaluation and Screening Criteria for the Brunswick 1 Shroud Indications to the Unit 1 Indications / Crack for BNP U1 Reactor Core Shroud Crack Investigation,"
Rev. 0 dated Nov 17,1993.
5.23 BWR Core Shroud Eva\\uation Report No. SL-4942 Rev. O, " Load Definition Guideline," dated November 11,1994.
5.24 Engineering Evaluation, ESR 97-00034," Evaluation of Unit 2 Core Shroud Indications and Operability Assessment," Rev 2 (B213R1)
Ccrolins Power & Light Company ESR 9700706 Brunswick Nuclear Plant - Unit 1 Revision 0 Engineering Evaluation Pag. 9 6.0 Decign Basis / Inputs e
a l H4 l l Shroud Parameter l
l H6A l H68 l l
Hreld H1 H5 Comments / References H7 lo.or,,..
~ P _, m,._.
lWekt Elev imm vessel toro (en )
388 375 317.375 219 250 183 250 179 250 127.188 o#P4s157 Re, o l shroud Mean Resue (in )
94 00 88 00 88.00 88.00 88.00 84.75 lshmud was TNriness (in )
1.500 1.500 1.500 1.500 1.500 1.500 Strese Leveve:
stas 42 n= o toes twnean cue nne
!NAl Primary Membrane (psa 250 234 210 617 594 603 sir per caacuwen lN/U Petmary nending fpsq 483 1006 1814 2162 2204 2946 os2t iot2. en o lErreagency Pm (psi) 257 244 221 629 606 617 l Emergency Pb (ps0 637 1340 2434 2905 2900 3989 lraufted Pnmary Membrane (pw) 818 844 821 1315 1292 1330 lFauned Pnmary Bendin0 (ps0 641 1375 2560 3110 3176 4432 Flow Stess (35m for 304ss) (pso 50700 50700 50700 50700 50700 50700 sm = 1e e km l Safety Factors:
Per(FSAR Sac 3 9 521 lNormawupset 2 250 2 250 2.250 2.250 2.250 2 250 eWRVIP41 Rm 2 l Emergency 1.500 1.500 1.500 1.500 1.500 1 500 reuned 1.125 1.125 1.125 1.125 1.125 1.125 lMethodologyinputs:
swRv1P41 Rn 2 linspecnon Uncertainty. Depth (vi) 0 11 N/A N/A N/A N/A N/A swRviP47 Foo. Se rack Granth Rate (inJhr )
5 00E45 5 00E45 5 00E-05 5.00E-05 5 00E 05 5.00E-05 m.o.cmon uncetarwy s, swmnP43 nWey critetta (2t) (In.)
30 3.0 3.0 3.0 3.0 30 su e na.s.4 e e. ce mecanon reeuti l0eys in Run Cyde 12 672 672 672 672 672 672 Pmune, can o-os enn cc are tecerta.wy l Hours h Run Cyde 12 16128 16128 16128 16128 16128 16128 lEv.-( a.s)
=0
=0 2.
2.
2.
2.
o
. s - r, -..,,,,,., -
lrtews toke s
.,...ws.
Y Y
Y Y
Y Y
kurpormed e ou emeram rivence Value et End of LWe (PW )
1.33E + 18 6.10E +20 2.34E+20 169E +18 9 25E+17 2.38E+ 14 gio s trPy to arous o l Acceptance Crtterta:
awmnP.ao (ou em,wn>
lMai Uncracked Ugament Lesic On )
159 106 190 215 215 224 slR 94 03 l Man Cract Depth nar 360 eaw on )
1.463 1 451 1 429 1.398 1 398 1.372 Pw BWRVIP41 Sacson 4 5
F Ccrolina Pow:r & Light Company ESR 9700706 Brunswick Nuclear Plant - Unit 1 Revision 0 Engineering Evaluation Page 10 7.0 Evcluation Dineral Conservatisms For all welds except HI Upper, assumed all areas uninspect.ed are fully cracked through-o wall and no credit was taken in analysis for these areas. (H1-lower,H4,HS,H6a,H6b,H7)
Used 5x 10' in/hr crack growth rate in lieu of measured or predicted BNP crack growth o
rates. (All welds)
Assumed all reported flaws were through wall. (H1-lower,H4,HS,H6a,H6b,H7) o UT consistently oversized the reported length on all flaws. (All welds) o Applied UT uncertainty for depth on all flaws where specific depth limit load analysis o
was run. (H1-upper)
Grew crack lengths (at both tips) for 3 cycles to take credit for 1 additional cycle of o
operation. Used 2000 days. Ac*ualplanned is 534 days of operation for fuel cycle 11, and 632 days for fuel cycle 12 and 705 days for fuel cycle 13 for a total of 1871 days.
Also assumed a 100% capacity factor.(H1-lower,H4,HS,H6a,H6b,H7)
Safetv Factors The analyses contained in this engineering evaluation used safety factors consistent with BSEP Up. dated FSAR Section 3.9.5.2.1. For loadings during normal operation (Level A) and upset (Level B) conditions, a factor of safety of 2.25 was used. For emergency (Level C) conditions, a facto. of safety of 1.50 was used. For faulted (Level D) conditions, a factor of safety of 1.125 was used.
Shroud Horizontal Weld Stresses Primary membrane (Pm) and Primary bending (Pb) stresses have beers updated to account for the latest design information as documented in Reference 5.21.
Ccrolins Power & Light Company ESR 9700706 Brunswick Nuclear Plant - Unit 1 Revision O Enginocring Evaluation Page 11 H1 Uoner Sidm This weld was inspected during the last outage. The 36 lug sets that exist on this dryer / separator support ring outside diameter preclude full inspection with the tooling currently available. Thus, only 18.8% of this weld is accessible. However, an excellent representative sample of the weld condition is obtained since the accessible 18.8% is uniformly distributed along the weld circumference.
Since al! cther horizontal core shroud welds were qualified beyond this refuel outage, H1 upper was the only weld that " fit" into the reinspection criteria of VIP-07 for this outage.
However, since a plant speclile analysis is acceptable and in accordance with the VIP documents, weld H1 will be evaluated for an additional cycle without reinspection this cutage. This approach was selected based on the following factors.
As documented in GENE-523-A107P-0794, separation at the H1 weld location will o
not impact any of the key functions required for response to a design basis accident.
Therefore, this weld is considered to be non-safety significant.
The cost of bringing in the inspection tooling required for this weld only, is substantial o
and would be better utilized to improve safety in other plant areas.
This weld experiences relatively small operating loads which requires a very small o
overall uniform ligament thickness (.037") to meet the FSAR safety factors.
The existing weld inspection results provides an excellent statistical representation of o
the weld due to its uniform coverage along the weld circumerence.
During the last outage the ID fillet weld was measured. This weld was not credited in o
the previous evaluatic but will be considered in this evaluation. A consistent fillet leg length was found to exist at the 28 inspected locations (see Attachment A).
Crack growth, while not precisely measured, is expected to be well below the currently used value of SE-5 inches per hour based on available BNP and industry core shroud re-inspection data.
Current analysis inouts summary:
This analysis starts with the raw data from the previous inspection, ref. 5.18. A review of the inspection data indicated a maximum measured crack depth of 0.66 inches. This value was used for the input crack depth in allinaccessible areas on the shroud outside diameter. In comparison, the average depth of the inspection data was 0.36", and twice the standard deviation,20, plus the average veas 0.36 + 0.28" = 0.64". Thus, the assumption of a 0.66" crack det + its uninspect i areas (81.2% of the circumference) is very conservative. No evidence of cracking was observed on the inside diameter of H1.
Of the 28 inspected locations from around the outside diameter, from both the upper and lower sides of H1. no indication of ID cracking was seen, if significant cracking on the ID existed, the probabili y of not finding it is very small given the sample size and uniform spacing characteristics of the inspection (see Attachment E for a graphical representation of the areas inspected). Therefore, no crack initiation or growth from the inside diameter was assumed in this analysis.
Ccroline Pow:r & Light Company ESR 9700706 Brur. wick Nuclear Plant - Unit 1 Revision 0 Engineering Evaluation Page 12 The inspection and measurement of the fillet weld on the ID of the shroud was performed last outage. The fillet weld leg along the support ring was consistently measured at 1.10 inches. This length was added to the shroud thickness of 1.5 inches to obtain the effective analysis thickness of 2.6 inches. This methodology is discussed in reference 5.6, Section 5.
A crack depth inpsection uncertainty of 0.106" was added to all crack depths. This uncertainty is specific to the UT technique used.
The crack growth of 5.0E-5 in/hr for cycle 11 and cycle 12 (1246 days) was added to all crack depths. The 1246 days is conservatively based on the planned cycle lengths,574 days for cycle 11 and 672 days for cycle 12.
H1 Lower, H4 H5, H6e, H6b, H7; As per the previous analysis, these welds were qualified for multiple cycles. They are rerun in this evaluation to incorporate revised design loads and the revised DLL program methodology of cracks taking compression.
8.0 Fluence
2 Wold H4 is the only weld expected to exceed the fluence threshold of 3E for load limit analysis. For that reason, the four (4) areas of H4 that exceed the threshold will be assumed fully cracked. These areas are conservatively taken to be azimuths 30 to 60, 120 to 150,210 to 240, and 300 to 330 degrees. These areas are conservative since the computed area is approximately 25 degrees per quadrant at pit.nt end of life (EOL).
See fig. 6 of ref. 5.24 for a graphical representation of fluence patterns (common for both units).
Carolins Powcr & Light Company ESR 9700706 Brunswick Nuclear Plant - Unit 1 Revision 0 Engineering Evaluation Page 13 Unit 1 Shroud ID Fluence Projections em i
a 500 r
f g
L.a J
l{
g
+ H3 2],
>W r
+ HS 3E20 Threshold g
s p --
0 10 11 12 13 14 15 16 17 le 19 20 21 cm. no As shown on the above graph, the design LEFM threshold will be crossed by H4 during the 12* cycle. No other welds are calculated to cross the threshold. Welds H1, H6a, H6b, & H7 are not shown on the chart as they are below the range of values graphed.
e
Ccrolina Pow:r & Light Company ESR 9700706 Brunswick Nuclear Plant - Unit 1 Revision 0 Engineering Evaluation Fage 14 9.0 Safety Factors:
The following table summarizes the minimum Unit 1 core shroud horizontal weld safety factors for the current evaluation.
Safety Factor Summary at End of Evaluation Period
~
W eld Normal Upset Emergency Faulted H1 Upper 19.76 16.18 9.65 H1 Lower 58.18 47.71 29.30 H4 21.29 16.68 11.78 HS 15.79 12.04 9.24 H6a 14.80 11.55 9.20 H6b 15.85 12.44 9.99 H7 14.00 10.79 8.69 4
e
Ccrolina Pow:r & Light Compent ESR 9700706 Brunswick Nuclear Plant - Unit
Revision 0 Engineering Evaluation Page 15 10.0 Reinspection Requirements:
Currently the NRC and the BWRVIP reinspection committee are resolving questions and concerns about the stated guidance of VIP-07.(Ref. 5.16). The, reinspection requirements for the Unit 1 shroud wolds willlist the Table 1 guidance of VIP-07 and the NRC suggested guidance which was issued in their SE.
BNP Unit 1 Core Shroud Weld Reinspection Data Weld Data H1U H1L H4 H5 H6a H6b H7
% insp. of total weld 18.8 79.8 77.6 76.8 76.6 77.8 76.8
% insp. found cracked 98.2 12.0 8.1 25.0 5.8 8.9 1.6 Faulted Stress Leve!
1.5 1.5 2.2 3.4 4.4 4.5 5.8 Date of Last insp.
10/96 10/96 10/96 10/96 10/96 10/96 10/96 Qual. per analysis 'i' 600 2000 2000 2000 2000 2000 2000 Interval per VIP-07 Spec.
8 yrs 8 yrs 6 yrs 8 yrs 8 yrs 8 yrs Date per VIP 07 Spec.
5"/04 10/04 10/02 10/04 10/04 10/04 Outage prior to Date Spec.
B115 B115 B114 B115 B115 B115 interval per NRC SE '2' Spec.
4/2 5/3 4/2 5/3 5/3 5/3 Outage prior to SE B113 B113 B113 B113 B113 B113 B113 Notes: 1. " Spec." refers to a plant specific analysis requirement.
- 2. The interval per the NRC SE is specified in years and cycles, i.e. 4 l 2 is 4 years or 2 cycles, whichever comes first.
- 3. " Qual." refers to the qualification time interval of the latest analysis in days.
ESR 9700706 Rev.
0 ATTACHMENT B Guideline for 10 CFR 50.59 Safety Evaluations ATTACHMENT 1 Page 1 of 5 10 CFR 50.59 Safaty Evaluation Screen Page 1 of L ACTMTV NO.
ESR 9700704 REV. 0 3
- 1.. OCES Titt ACTMTY REQUIRE A CHANGE TO THE OPERATINO UCENSE OR TECHNICAL $PECIFK:ATION7
[]YES [X)NO Seele: Thle document evaluates the as found condition of the core s'hroud hortrontat welde for Unit i based on NDE inspections perbemed dunno the Bit tR1 outaae AN UFVR safety factore have been esbened for the core shroud weide for the 12* run cycle No envyrgt, 3_ent modMcations or operational restrtettone are reaulred as e result of thle ESR. Therefore thle ESR does not reautre e chance to 8
ts_operemna licente or enicet spec,nceoon Note: If Yes, and the scope of the activity le Smited to a Technical Specincation/ Operating Ucensing change, then complete Section M i
of thee wrm, and procese por plant procedure, if the scope of the activity le not limited to a Tech. Spec. or OL change,in addition to processing a Tech. Spec. of OL cfkage ratueet, continue the screening process, if No, continue the screening procese, 2.
IS THE ACTMTV FULLY BOUNDED BY A PftEVIOUSLY PERFORMED 10 CFR 50.55 SAFETY EVALUATION 7
[a)YES []NO Evelm No, The moet recent 10CFR 50.69 Safety Evaluationlgontained in EER S40077, The previous plant speelfic evolustions were referenced to one fuel cycle The conclusions from ESR 9700708 show that the required UF SAR safety margine are malatained for et least one addRional fueJIcle (cycle 171. Therefnre. the previous safety evaluatione are ccmsidered boundina. Additionaire a eeneric eafety assosoment of shroud weld failuree wee provided try the BWROO in 1994 *8wR Shroud Crackina Generic Safety Aeassament', GEN 2-823 A107P4764 Rev6elon 1, Aucuet 1994. A DNP plant spectftc assetement wee prov6ded in response to Generic Letter 9443.
Note:If Yes, attach a copy or prov6de document number for retrieval capability of the previousty performed 10 CFR 80.59 Safety Evolvation and complete Section 8 of thle form, if No, continue the screening procese.
3.
DOES THE ACTMTY MAKE CHANGES TO THE FACILITY AS DESCRIBED IN THE SAR7
[]YES [JNO J
Baste:
1 Uet SAR ltems/ Sections reviewed:
4 DOES THE ACTMTV MAKE CHANGES TO PROCEDURES AS DESCRIBED IN THE SAR7
[]YES []NO Baste:
Uet SAR leema/Secitone reviewed:
1 S.
DOES Titt ACTMTY INVOLVE A TEST OR EXPERIMENT NOT DESCRIBED IN THE SAR7
[ ) YES []NO Beele:
Ust SAR ltems/Sectione revlowed:
NOTE:If any question 3 through 5 le oneworod YES,then mart $4 ction e6 Not ApplV able (NIA) and complete Unreviewed Safety Question Determinetton, otherwise complete Sociton 86.
4.
DISCIPLINE PRINT NAME SIGNATURE tot QSR:
Structuret i
L. Yemma f A W
Date: 01/22/98 Other QSR:
1-t Date:
Othe, CSR:
/
f
, r-Date:
2nd QSR:
Mechanical i
M OHver i
L Date: 7 8 19 Attach addluonal sheets if needed.
10 CFR 50.59 Rev.5 Page 13 of 37 OAl-109 Rev.9 Page 25 of 55
f l
-ESR 9700706 Rev.
0 Page /7 ATTACHMENT C Page 1 of 2 Item Classification Form DOCUMENT NO.
ESR 9700706 REV.NO.
O yl3 NO Does this item geludingporgchanges to the faci.lity or
[]
[X]
1.
. NN? fSee Section 5.0 of 0Al-109 obstr0c$iok.
2.
Does this item involve a chanae to the Off-Site Dose Calculation
- Manual? (if yes, see Section3.0 of 0Al-109 for instruction.)
[]
[X]
3.
Does this item constitute a chanae to the Process Control Program? (if yes, see Section 5'.0 of 0Al-109 for instruction.)
[]
[X) 4.
Does this item involve a major chance to a Radwaste Treatment System? (if yes, see Section 5.0 of'0Al-109 for instruction.)
[]
[X]
5.
2oes this item invo ve a change to the Technical Specificatiort ON N.Y or fhher i'nSt ucTone Secki SYoNAU-1 environmental. impact the NPDES Permit or a "significant Does this item
- 6.
Attachment A to make th(is determination, if yes, see Section 5.0[]
[X]
impact"? Use t1e guidance contained in of 0Al-109 for instruction.)
7.
Does this item involve a change to a previously accepted:
I.
e0ty an nc$udkTr$ining, Qualification, and ontingency Ians IX mergencyPlan? )?
[X.
c.
(if yes see Section 5.0 of 0Al-109 for instruction and complete a 10 CFh 50.54 evaluation using Attachment D.)
8.
Does the item involve a chance to the following Fire Protection Proaram Documenis: OPLP-01; OPLP-01.1;
[]
[X]
. OPLP-01.2: Section 5.0 of 0Al-109 for instruction.)
or DPLP-01.5?
(liyes, see OAl-109 Rev.9 Page 50 of 54
ESR 9700706 Rev.
0 Page /8
- ATTACHMENT C Page 2 of 2 Item Classification Form DOCUMENT NO.
ESR 9700706 REV.NO.
0 YES NO Specification, iriclud. quire a revision to the Imoroved Technical.)
[xj Does this chanae re r
1.
Requirements Manual?associat-e bases or the Technical ino (If yes, see Section 5.0 of 0Al-109 for instruction.)
References:
' identify specific references used for any "Yes" answer.
0Al-109 Rev. 9 Page 51 of 54
i ESR 97o 0 ?% Rev o Anichmee A PnI CAROLINA POWER & LIGHT COMPANY Report No. N/A NDE REPORT /
SUMMARY
Page 1
of 2 Plant / Unit Component (s) or item (s) Examined Brunswick Unit 1 RPV SHROUD H-1 WELD Procedure (s) Used: N/A The purpose of this report is to document the actual profile measurements of the ID surface fillet weld on the shroud to ring weld (H 1). These measurements were made using the look-down 45' shear wave transducer scans that were performed during the 1996 fall refueling outage shroud inspection. The inspection was performed in accordance with GE procedure GE BRU 503V4 (PROCEDURE FOR THE AUTOMATED ULTRASONIC EXAMINATION OF THE SHROUD ASSEMBLY WELDS). The automated data collection system used was the GE Smart 2000 system.
Measurements wue made in all areas scanned from above weld H 1 (A total of 28 Lug sets: 2-3 thru 16-17 and 21-22 thru 33-34). The dimensions shown on attached page 2 represent the profiles at all 28 locations.
'Ihe distances along the ring OD surface represent the sound exit point from reference 0" (surface distance) and the distances along the 45' lines represent the sound metal path (MP) to the fac: of the fillet. The fillet surfaces between the lines were interpoleted to complete the profile. Due to beam spread and the low amplitude of the signal seen at Se deepest measurement (2" surface and 3.9" MP), the fillet profile was not
~
extended out to the plotted point on page 2. All of the other reflectors were high amplitude signals indicating intersection with a near perpendicular reflector. Ne evidence of cracking was seen in the upper ID toe area of the fillet (as already reported by GE).
As shown on page 2, the horizontal leg length of the ID fillet weld is 1.1 inches in length at all areas examined.
Reported By Level Date dQLg UT L-Ill 6 November,1996 Reviewed By Title Date CPL NDE 30, Rev. O,03/94
8 3
3 Y
J 5
k
- -aT !
R
- o o
ei 1
.e
="
u e
.~
E K 3 r,
da s
f
)(
//
\\
.x y a a 4R 4.
9 3
+
v'
,y s._
-- d a:
'd.
~
xw s
k.4 M.
-kg j
5 1
.W
)
3 0
Y am e Zk V -t"T"HY o N x 2.c o L 4 v s :f
_ _ _ _ - __