ML20092C845
| ML20092C845 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 12/31/1994 |
| From: | Henrie D, Muse J, Ranganath S GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20091F935 | List: |
| References | |
| FOIA-95-188 GE-NE-523-A181, GE-NE-523-A181-1294, NUDOCS 9509130159 | |
| Download: ML20092C845 (12) | |
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Revision 0 Class 11
-i G GE Nuclear Enern DRF Al2 00061 3ENERALELECTRIC COMPAN FROPRIETARY INFORMATION l
Commonwealth Edison Comany Dresden Nuclear Power Station, Units 2 & 3 Primary Structure Seismic Models i
December 1994 l
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Prepared By:
D.K.Henrie, Project Manager l
Quad Cities 1 & 2, Seismic Model Generation i5 m
VeriGed Byh.A. Muse, Eng'ineer i
Engineering and Licensing Consulting Services i
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Approved By:
S. Ranganath, Profects Manager Engineering and Licensing Consulting Services i
I Dresden Units 2 and 3 - Primary Structure Seismic Models g 91 g 9 950830 IRWIN95-188 PDR
Revisiern u Class 11
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DRF Al2 00061 GE Nuclear Enert*v NTRALELECTRIC COMPO rRUPRIETARY INFORMAT PROPRIETARY INFORMATION NOTICE This document contains proprietary information of the General Electric Co and isfurmshed to Commonwealth Edison Company, in conpdence solel or purposes stated in the purchase order between Commonwealth GE. Commonwealth Edison Company shallnotpublish or athenvise disclose thi document or the niformation to others without the written consent of GE except as provided in such purchase order andshall return this document at the IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The vuly undertaking of the General Elecinc Company (GE) respecting in h Edison this document are containedin the purchase order between Commomvealt Company and GE. and nothing contained in this document shall be consu changmg the purchase order. The use of this information by anyone other th Commonwealth Edison Company, orfor anypurpose other than thatfor which it miended under such purchase order in not authorized; and with respect to any miauthori:ed use GE makes no representation or warranty, and assumes no lia to the completeness, accuracy, or usefulness of the information contained in document, or that its use may not infringe privately owned rights.
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i Page i Dresden Units 2 and 3 - Primary Struaure Seismic Modets
GE-NE-323-Al31-1294 Revision 0 Class 11 DRF A12 00061 GE Nuclear Enertov I
i Table of Contents J
Pare No, Section 1
1.0 Introduction l
2
2.0 Purpose and Scope
3, 2
3.0 Seismic Models 4
4.0 Baseline Model Analysis & Results 1
6 5.0 References I
.i Al Appendix A - Figures B1 1
Appendix B - Tables Cl Appendix C - SAPG07 Input Geometry Listings for N-S and E-W Seismic Models GPIERALELECTRICCOMP rMRIETARY INFORMATI i
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k Dresden Units 2 and 3 - Primary Structure Seismic Models Pageii l
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1.0 INTRODUCTION
The licensing basis seismic design adequacy evaluations for the Dres Station, Units 2 and 3 primary structures were completed in the la early nineteen seventies. The primary structure seismic models evaluations are referred to as the " benchmark" seismic models in this discussed in Section 3.0 below. The seismic models reconstructed fo i
d l in this the benchmark models, are referred to as " regenerated baseline" seism c mo report.
In recent years, both the NRC staff and the nuclear industry hav i
W ater issue of concern the Intergranular Stress Corrosion Cracking (IGSCC)
Reactor (BWR) internal components, Reference 5.7. The core shr be prominent among the internal components susceptible to IGSCC. S h
inspections at Quad Cities Unit I and Dresden Unit 3 revealed sig core shroud circumferential welds. Due to the 360* extent of the crac i
d elevations where extensive cracking had not been previously observed, t d
Unit related analyses (Refereces 5.8 and 5.9) performed in the Spring of 199 3 and Quad Cities Unit I were especially noteworthy.
The analytical evaluations corresponding to and the reactor brought to a safe shut-down condition regardless of f
cracking at the core shroud circumferential welds. This allowed ld reactor untill the next planned fuel outage, at which time appropriate be most favorably implemented.
Conceptually, the shroud repair to be impleme d
comprised of: (i) two sets oflinearly clastic, la core support plate elevation) in conjunction with (ii) a set of ve h
upper shroud flange and the lower shroud support plate which st shroud to the RPV.
In order to generate loads for the design adequacy evaluation of the I
hardware, it is necessary to reconstruct the primary structure seis l
licensing basis seismic design adequacy evalu cracking at the shroud circumferential welds will significantly Page1 Dresden Units 2 and 3 - Primary Structure Seismic Models
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GENE Letter, M. D. Potter - GE Shroud Project Engineer to Kenneth Hutko -
Comed Shroud Project Engineer, Performance impact of shroud repair leakage for Dresden Units 2 & 3, dated May 18,1995 (B13-01749, MDP-9536) i I
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e GENuclearEnergy CeneretEwctnc Company May 18,1995 175 Curtner Awe, SunJose. CA 95125 cc:
R. Svarney E. R. Mohtashemi B13-01749 I
MDP-9536 To:
Kenneth Hutko Comed Shroud Project Engineer i
From:
M. D. Potter W
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GE Shroud Project Engineer
SUBJECT:
PERFORMANCE IMPACT OF SHROUD REPAIR LEAKAGE FOR DRESDEN UNITS 2 AND 3 i
Reference:
DRFNo. B13 01749.
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- 1. Introduction i
The hardware designed to repair the shroud with identified cracks for Dresden Units 2 and 3 1
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machining of eight holes through the shroud support plate. Each of these holes will have some(
i which will allow leakage flow to bypass the steam separation system. In addition, potential {
the weld cracks (HI through H8)and the replacement access hole cover is also considered his letter j
reports the leakage flow for 100% rated power and core flow.
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- 2. Evaluation s
4 2.1 Leakage Flow Evaluation
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De most restrictive flow area for leakage through the holes in the shroud support plate is based on conservative gap between the adjacent surfaces of the shroud support plate and the lower supporj
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In addition, there are a total of eight circumferential shroud welds (H1 - H8) that are con j
leakage paths - two above the top guide support ring, three on the upper shroud betw ring and the top guide support ring, and three on the su wer shroud below the core support conservatively assumed that each of these welds develops a corr.plete circumferential crack t 0.001 inches.
- De leakage flows for 100% rated power and core flow are summanzad in Table 1. Dese 2
are based on applicable loss coefficients and reactor internal pressurs differences (RIPD's) across 4
l applicable shroud components. De replacement access hole cover leakage is based on informa referenced DRF. Leakage from the weld cracks above the top guide support ring is assumed to i
phase fluid at the core exit quality. leakage from the remaining paths below the top gui
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considered single-phase liquid. All of the leakage flows bypass the steam separators and dryej leakage flows below the shroud support nng also bypass the core. De results show that from the repair holes, weld cracks and the access hole cover result in a combined leakage o core flow.
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Table 1. Saanmary ofI4akage Flows at Rated Power and Flow 4
Leakage flow (gpm)
Shroud head flange pockets 1600 Weld cracks 140 Repair holes in support plate 325 Access hole covers 180 Leakage-to. core Mass flow (%)
Shroud head flange pockets 0.21 Weld cracks 0.04 i
Repair holes in support plate 0.12 Access hole covers 0.07 he steam portion of the leakage flows will contribute to increasing the total carryunder from the steam separators. De impacts of the total leakage on the steam separation system performance, jet pump performance, core monitoring, fuel thermal margin, emergency core cooling system (ECCS) performance i
and fuel cycle length are evaluated a summarized in the following subsections.
2.2 Steam Separation System he leakage flow through weld cracks H1 and H2 occurs above the top guide support ring and includes steam flow, which effectively increases the total carryunder in the downcomer by about 0.03% at rated i
conditions. De carryunder from the separators is based on the applicable separator test data at the lower limit of the operating water level range. The combined effective carryunder from the separators and the shroud head leakage is about 0.18% and is bounded by the design value.
2.3 Jet Pumps he increased total carryunder will decrease the subcooling of the flow in the downcomer. This in turn reduces the margin tojet pump cavitation. However, because the total carryunder meets the design-condition carryunder value, there is no impact on jet pump performance compared with the design
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condition.
j 2.4 Core Monitoring The impact of the leakage results in an overprediction of core flow by about 0.21% of core flow. His 1
overprediction is small compared with the core flow measurement uncertainty of 2.5% forjet pump plant used in the MCPR Safety Limit evaluations. Additionally, the decrease in core flow resulting from the overprediction results in only a 0.1% decrease in calculated MCPR. Derefore, it is concluded that the impact is not significant.
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2.5 Anticipated AbnormalTransients De code used to evaluate performance under anticipated abnormal transients and determine fuel thermal margin includes carryunder as one of the inputs. De effect of the increased carryunder due to leaka results in greater compressibility of the downcomer region and, hence, a reduced maximum vessel i
Since this is a favorable effect, the thermal limits are not impacted.
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i 2.6 Emergency Core Cooling System leakage through weld cracks H1 and H2 results in slightly increased w.p..ds that causes the initial core inlet enthalpy to increase slightly, with a corresponding decrease in the core inlet subcooling. However, W== the total downcomer carryunder still meets the design value, there is no impact on the emergency core cooling system (ECCS) performance from this effect compared with the design conditions. Another effect of the leakage flows from the repair holes and the weld cracks is to decrease the time to core uncovery slightly and, also to increase the time that the core is uncovered. he combined effect has been assessed to increase the peak cladding : # w (PCD for the limiting LOCA event by less than 30*F.
De current analysis basis yields a LOCA PCT of about 2045'F for the design basis LOCA with LPCI injection failure. De 10CFR50.46 regulatory limit PCT is 2200*F. hc== the maximum potential effect on the design basis LOCA PCTis very small, there is no adverse effect on the margin of safety. His impact is sufficiently small to be judged insignificant, and hence, the licensing basis PCT for the normal condition with no shroud leakage is applicable. De sequence of events remains essentially unchanged for the LOCA events with the shroud head leakage.
2.7 Fuel Cycle Length De increased carryunder due to leakage flow above the top guide support ring results in a slight increase in the core inlet enthalpy, compared with the no-leakage condition. The combined impact of the reduced core inlet subcooling aixt the reduced core flow due to the leakage results in a t.tinor effect (-0.8 days) on fuel cycle length and is considered negligible.
- 3. C%
De impact of the leakage flows through the shroud repair holes and the potential weld cracks in the shroud have been evaluated. De results show that at rated power and core flow, the leakage flows from the repair i
holes and the weld cracks are predicted equal to a combined leakage of about 0.44% of core flow (including potential replacement access hole cover leakage). These leakage flows are sufficiently small so that the steam separation system performance, jet pump petformance, core monitoring, fuel thermal margin and fuel cycle length remain adequate. Also, the impact on ECCS performance is sufficiently small to bejudged insignificant, and hence. the licensing basis PCT for the normal condition with no shroud leakage is applicable.
M. D. Potter i
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