ML20086G318
| ML20086G318 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/31/1991 |
| From: | Biglieri N, Deaver G, Nghiem H GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20086G317 | List: |
| References | |
| RDE-41-0791, RDE-41-791, NUDOCS 9112050039 | |
| Download: ML20086G318 (11) | |
Text
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GE NUCLEAR ENERGY RDE 41-0791 SAN JOSE, CALIFORNIA DRF B11 00527 JULY 1991 CORE SPRAY LINE CRACK ANALYSIS UPDATE FOR BRUNSWICK STEAM ELECTRIC PLANT 1
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SAFETY EVALUATION JULY 1991 Prepared by:
K.R. Kotak Reviewed by:
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H.X. Nghtem Reactor t.omponent Design Approved by:
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G.A. Deaver, Manager Reactor Component Design Approved by:
b N.4 Biglieri, Manager Reactor Design Engineering
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GE NUCLEAR ENERGY RDE.41- 0791 SAN JOSE, CALIFORNIA DRF B11 00527 JULY 1991 1HPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of the General Electric Coupany respecting information in this document are contained in the contract between Carolina Power & Light Company and General Electric Company, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Carolina Power & Light Company, or for any purpose other than that for which it is intended under such contract is not authorized; and eith respect to any unauthorized use, General Electric Company makes no representation or warranty, and assumes no li.bility as to the completeness, accuracy or usefulness of the information contained in this document.
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GE NUCLEAR ENERGY RDE-41 0791 SAN JOSE, CAllf0RNIA DRF B11 00527 JULY 1991 TABLE OF CONTENTS f5[
4 1.
INTRODUCTION AND BACKGROUND 1
1.1 History 1
1.2 Core Spray Line Hodification Design 1
2.
REVIEW Of PREVIOUS ANALYSIS 2
2.1 Applicability of Previous Analysis 2
2.1.1 Crack Leakage Estimate Review 3
2.1.2 Core Spray Pipe Structural 3
Integrity Review 2.1.3 Lost Part Analysis I,cview 4
2.1.4 Loss-of Coolant Accident 4
Analysis Review 2.2 Summary of Review 5
3.
SAFETY EVALVATION CONCLUSION REVIEW 5
REFERENCES 6
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GE NUCLEAR ENERGY RDE 41 0791 SAN JOSE, Calif 0RNIA DRf B11 00527 JULY 1991
1.0 INTRODUCTION AND BACKGROUND
1.1 History During the refueling and maintenance outage of Unit 2 following fuel cycle 8. CP&L completed NOT (UT & PT) erimination of the core spray lines on the piping adjacent to both sides of the tec-box on both the north and south core spray line headers.
These examinations were performed in addition to those required by inspection and Enforcement (IE) Bulletin 80 13 and those suggested by GE Nuclear Energy Service Information Letter (Sil) Number 289.
Olmensions of indications associated with IGSCC on the right side of the north tee box piping (see fig >>re 1) were reported as follows:
1)
The indication is approximately 4.90 inches in length along the inside of the pipe (111 degrees of the inner circumference),and 2) approximately 1.90 inches in length along the outside of the pipe (39 degrees of the outer circumference).
The evaluation performed per Reference 3 determined the acceptability of continued operation without any repair for fuel cycle 9.
1.2 Core Snray line Modification Design During the refueling outage that will follow fuel cycle 9, a modification design will be implemented consisting of two bracket assemblies, the first to cover the cracked tee box location at 90' azimuth location and a second for the other tee-box at 270' azimuth location. Each bracket assembly consists of an upper and -.
GE HUCLEAR ENERGY RDE-41-0791 SAN JOSE, CALIFORNIA DRf B11 00527 JULY 1991 lower bracket welded across the piping arms and tre box (see figure 2), Bracket assemblies will assure that the struttural integrity of piping is maintained.
Each bracket assembly will be structurally adequate to hold the tee box and pipe headers together in the event that cracks on bcth sides of tee box propagate through the entire pipe cross section, Desian Cri*.oria of Brackets:
Material Selection The brackets will be made of low carbon (<.02Y.) austenitic i
stainless steel material in annealed condition which is resistant to IGSCC in creviced condition.
11ress limits The allowable stress in the brackets will be established using the ASME boiler and pressure vessel code, Section !!! as a guide.
The bracket is not classified as an ASME code component.
- However, using the ASHE code design limits will ensure structural adequacy of the hardware.
2.0 PREVIOUS ANALYSIS APPLICABILITY REVIEW 2,1 ApolicabilitY of Previous Analysis The evaluation performed per Reference 3 to determine the acceptability of continued operation for fuel cycle 9 were reviewed to determine which areas might require updating due to.
GE NUCLEAR ENERGY RDE 41-0791 SAN JOSE CALIFORNIA DRF Bll 00527 4
JULY 1991 installation of brackets and to extend the current evalva ion to support operation in the upcoming cycles.
The results of this review are documented below:
2.1.1 Crack Leakaoe Estimate Revity The core spray line crack leakage rates utilized in previous evaluation discussed in Section 2 of References 1 and 2 indicated that crack growth would be small leading to a virtual arrest condition prior to reaching about 180' of the piping circumference.
The leakage for a 180' crack in the Brunswick Unit 2 core spray line was and is still conservatively estimated to be 20 GPH.
2.1.2 Core Snray Pine Structural Intearity Revi w t
The basic conclusions of the structural integrity evalu; tion documented in Section 3 of the Reference 1 report are still valid and are still part of the summary provided below:
1.
The driving force for crack extension is expected to be small when the crack length reaches 180* of the piping circumference. Therefore, a slowdown or a virtual crack arrest condition is expected at this point.
2.
A through wall crack of up to 235' of the piping circumference without the bracket assembly can be tolerated without gross failure of the core spray line..
GE HUCLEAR ENERGY RDE 41 0793 SAN JOSE, CAllFORNIA DRF B11 00527 JULY 1991 3.
To assure that the piping can accommodate any of the postulated load:,, two bracket assemblies (90' and l
270') will be welded across the piping arms and tee box. These bracket assen.blics will assure that the structural integrity of the piping is maintained even if all four weld joints (i.e., two on each side of each tee box) develop cracks that propagate through the entire pipe cross section.
2.1.3 Lost Part Analysis Review Although it is anticipated that the piping will not break, particularly with the brackets installed, an existing lost part analysis documented in Section 4 of Reference I was re evaluated to account for additional lost pieces of the brackets and also a whole bracket as a lost piece.
It is concluded that no safety concerns are posed by any such postulated lost parts. The probability of unacceptable corrosion or other chemical action is not changed from the present condition since both piping and bracket material are selected for the reactor vessel environment.
2.1.4 Loss of Coolant Accident Analysis Review Loss of coolant analysis documented in Section 2.1.4 of Reference 3 is independent of bt acket installation and therefore still valid.
It was noted in Section 3.0 of Reference 3 that one 180' arc length crack resulted in bounding loss of core spray flow of 20 GPM and LOCA analysis was performed adequately with an assumed reduction in the delivered flow rate of 725 GPM from the designed pump flow rate. Calculations show that even if all four 4
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GE NUCLEAR ENERGY RDE-41-0791 g
SAN JOSE CAtlFORNIA DRF B11-00527 JULY 1991
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,g l weld o ints develop a crack through the entire pipe cross d
section, the total leakage is still within the margin of i
the core spray system and therefore, the conclusions of N
'ef. 3 are still valid.
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J "7i The nec2ss' cy updates as applicable are documented in the previous sections of this report. The updated evaluations are consistent with the previous conclusions and support continued operation of Brunswick Unit ? for additional cycles.
3, SAFETY EVALUATION CONCLUSION REVIEW Based on the evaluations presented in References 1 and 3 and as updated herein, it is concluded that continued operation with the core spray line indications and the addition of brackets for additional cycles does not constitute an unreviewed safety queuion or a significant safety hazard. The reasons for the safety evciustion conclusions provided in Section 3 of Reference 3 are :.
i valid and do nat require updatino.
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GE HUCLEAR ENERGY RDE-41 0791 SAN JOSE, CAllFORNIA DRF B11-00527 Jt!LY 1991 REFERENCES 1.0 "Cora Spray line Crack Analysis for Brunswick Steam Electric Plant Unit 2", General _ Electric Company, EAS-14 0388, F. arch 1988.
2.0
" Brunswick Steam Electric Plant Units 1 and 2 SAFER /GESTR LOCA Loss of Coolant Acc it Analysis", GE Nuclear Energy, NEDC-31624P, Class III, September 196d.
3.0
" Core Spray Line Crack Growth Analysis Update for Brunswick Steam Electric Plant Unit 2," General Electric Company EAS-03 0190 (Supplement 1 of EAS 14-0388, March 1988) January, 1990. l l
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